ML11308A097
| ML11308A097 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 10/28/2011 |
| From: | Entergy Operations |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| GNRO-2011/00093 | |
| Download: ML11308A097 (124) | |
Text
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-i APPENDIX B AGING MANAGEMENT PROGRAMS AND ACTIVITIES TABLE OF CONTENTS B.0 INTRODUCTION................................................ B-1 B.0.1 Overview................................................... B-1 B.0.2 Format of Presentation......................................... B-1 B.0.3 Corrective Actions, Confirmation Process and Administrative Controls.... B-2 B.0.4 Operating Experience.......................................... B-3 B.0.5 Aging Management Programs................................... B-3 B.0.6 Correlation with NUREG-1801 Aging Management Programs.......... B-6 B.1 AGING MANAGEMENT PROGRAMS AND ACTIVITIES...................... B-14 B.1.1 115 KV Inaccessible Transmission Cable..........................B-14 B.1.2 Aboveground Metallic Tanks....................................B-18 B.1.3 Bolting Integrity...............................................B-20 B.1.4 Boraflex Monitoring...........................................B-23 B.1.5 Buried Piping and Tanks Inspection...............................B-25 B.1.6 BWR CRD Return Line Nozzle...................................B-27 B.1.7 BWR Feedwater Nozzle........................................B-29 B.1.8 BWR Penetrations............................................B-31 B.1.9 BWR Stress Corrosion Cracking.................................B-32 B.1.10 BWR Vessel ID Attachment Welds...............................B-34 B.1.11 BWR Vessel Internals.........................................B-36 B.1.12 Compressed Air Monitoring.....................................B-40 B.1.13 Containment Inservice Inspection - IWE...........................B-42 B.1.14 Containment Inservice Inspection - IWL...........................B-44 B.1.15 Containment Leak Rate........................................B-46 B.1.16 Diesel Fuel Monitoring.........................................B-48 B.1.17 Environmental Qualification (EQ) of Electric Components..............B-51
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-ii B.1.18 External Surfaces Monitoring....................................B-54 B.1.19 Fatigue Monitoring............................................B-56 B.1.20 Fire Protection...............................................B-59 B.1.21 Fire Water System............................................B-62 B.1.22 Flow-Accelerated Corrosion.....................................B-66 B.1.23 Inservice Inspection...........................................B-68 B.1.24 Inservice Inspection - IWF......................................B-70 B.1.25 Inspection of Overhead Heavy Load and Light Load (Related to Refueling)
Handling Systems............................................B-73 B.1.26 Internal Surfaces in Miscellaneous Piping and Ducting Components.....B-75 B.1.27 Masonry Wall................................................B-77 B.1.28 Non-EQ Cable Connections.....................................B-79 B.1.29 Non-EQ Inaccessible Power Cables (400 V to 35 kV).................B-81 B.1.30 Non-EQ Instrumentation Circuits Test Review.......................B-85 B.1.31 Non-EQ Insulated Cables and Connections.........................B-87 B.1.32 Oil Analysis..................................................B-89 B.1.33 One-Time Inspection..........................................B-91 B.1.34 One-Time Inspection - Small-Bore Piping..........................B-93 B.1.35 Periodic Surveillance and Preventive Maintenance...................B-95 B.1.36 Protective Coating Monitoring and Maintenance.....................B-99 B.1.37 Reactor Head Closure Studs....................................B-101 B.1.38 Reactor Vessel Surveillance....................................B-103 B.1.39 RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants................................................B-105 B.1.40 Selective Leaching............................................B-107 B.1.41 Service Water Integrity.........................................B-109 B.1.42 Structures Monitoring..........................................B-111 B.1.43 Water Chemistry Control - BWR.................................B-116 B.1.44 Water Chemistry Control - Closed Treated Water Systems............B-118 B.2 REFERENCES................................................. B-122
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-1 B.0 INTRODUCTION B.0.1 OVERVIEW The aging management review results for the integrated plant assessment of Grand Gulf Nuclear Station (GGNS) are presented in Sections 3.1 through 3.6 of this application. The programs credited in the integrated plant assessment for managing aging effects are described in this appendix.
Each aging management program described in this appendix has ten elements in accordance with the guidance in NUREG-1800 (Reference B.2-1) Appendix A.1, Aging Management Review
- Generic, Table A.1-1, Elements of an Aging Management Program for License Renewal. For aging management programs that are comparable to the programs described in Sections X and XI of NUREG-1801 (Reference B.2-2), Generic Aging Lessons Learned (GALL) Report, the ten elements have been compared to the elements of the NUREG-1801 program. For plant-specific programs that do not correlate with NUREG-1801, the ten elements are addressed in the program description.
B.0.2 FORMAT OF PRESENTATION For those aging management programs that are comparable to the programs described in Sections X and XI of NUREG-1801, the program discussion is presented in the following format.
Program
Description:
abstract of the overall program.
NUREG-1801 Consistency: summary of the degree of consistency between the GGNS program and the corresponding NUREG-1801 program, when applicable (i.e., degree of similarity, etc.).
Exceptions to NUREG-1801: exceptions to the NUREG-1801 program, including a justification for the exceptions (when applicable).
Enhancements: future program enhancements with a proposed schedule for their completion (when applicable).
Operating Experience: discussion of operating experience information specific to the program.
==
Conclusion:==
statement of reasonable assurance that the program is effective, or will be effective, once implemented with necessary enhancements.
For plant-specific programs, a complete discussion of the ten elements of NUREG-1800 Table A.1-1 is provided.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-2 B.0.3 CORRECTIVE ACTIONS, CONFIRMATION PROCESS AND ADMINISTRATIVE CONTROLS Three elements common to all aging management programs are corrective actions, confirmation process and administrative controls. Discussion of these elements is presented below.
Corrective actions have program-specific details which are included in the descriptions of the individual programs in this report, but further discussion of the confirmation process and administrative controls is not necessary and is not included in the descriptions of the individual programs.
Corrective Actions GGNS quality assurance (QA) procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR Part 50, Appendix B.
Conditions adverse to quality, such as failures, malfunctions, deviations, defective material and equipment, and nonconformances, are promptly identified and corrected. In the case of significant conditions adverse to quality, measures are implemented to ensure that the cause of the nonconformance is determined and that corrective action is taken to preclude recurrence. In addition, the root cause of the significant condition adverse to quality and the corrective action implemented are documented and reported to appropriate levels of management. The corrective action controls of the GGNS (10 CFR Part 50, Appendix B) Quality Assurance Program are applicable to all aging management programs and activities during the period of extended operation.
Confirmation Process GGNS QA procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR Part 50, Appendix B. The GGNS Quality Assurance Program applies to GGNS safety-related structures and components.
Corrective actions and administrative (document) control for both safety-related and nonsafety-related structures and components are accomplished in accordance with the established GGNS Corrective Action Program (CAP) and Document Control Program. The confirmation process is part of the CAP and includes the following:
Reviews to assure that corrective actions are adequate.
Tracking and reporting of open corrective actions.
Review of corrective action effectiveness.
Any follow-up inspection required by the confirmation process is documented in accordance with the CAP. The CAP constitutes the confirmation process for aging management programs and activities. The GGNS confirmation process is consistent with NUREG-1801.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-3 Administrative Controls GGNS QA procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR Part 50, Appendix B. The GGNS Quality Assurance Program applies to GGNS safety-related structures and components.
Administrative (document) control for both safety-related and nonsafety-related structures and components is accomplished per the existing document control program. The GGNS administrative controls are consistent with NUREG-1801.
B.0.4 OPERATING EXPERIENCE Operating experience for the programs and activities credited with managing the effects of aging was reviewed. The operating experience review included a review of corrective actions resulting in program enhancements. For inspection programs, reports of recent inspections, examinations, or tests were reviewed to determine if aging effects have been identified on applicable components. For monitoring programs, reports of sample results were reviewed to determine if parameters are being maintained as required by the program. Also, program owners contributed evidence of program success or weakness and identified applicable self-assessments, QA audits, peer evaluations, and NRC reviews.
Operating experience is used at GGNS to enhance plant aging management programs. External nuclear industry operating experience is screened, evaluated, and acted on to prevent or mitigate the consequences of similar age-related degradation. External operating experience may include NRC generic communications (e.g., Generic Letters, Bulletins, Information Notices) and other documents (e.g., 10 CFR 21 Reports, Licensee Event Reports, Nonconformance Reports). Internal operating experience may include such things as event investigations, trending reports, lessons learned from in-house events, self-assessments, and the 10 CFR 50 Appendix B corrective action process.
Site procedures for the evaluation of these sources of operating experience remain in place as the site continues operation through the license renewal period. These procedures implement two existing programs that monitor, on an ongoing basis, industry and plant-specific operating experience that includes, but is not limited to, future operating experience related to the effects of aging on in-scope structures and components. These programs are the Operating Experience Program and the CAP. The evaluations completed under these two programs ensure that aging management programs continue to be effective in managing the aging effects for which they are credited.
B.0.5 AGING MANAGEMENT PROGRAMS Table B-1 lists the aging management programs described in this appendix. Programs are identified as either existing or new. The programs are either comparable to programs described in NUREG-1801 or are plant-specific. The correlation between NUREG-1801 programs and GGNS programs is shown in Table B-2.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-4 Table B-1 Aging Management Programs Program Section New or Existing 115 kV Inaccessible Transmission Cable B.1.1 new Aboveground Metallic Tanks B.1.2 new Bolting Integrity B.1.3 existing Boraflex Monitoring B.1.4 existing Buried Piping and Tanks Inspection B.1.5 new BWR CRD Return Line Nozzle B.1.6 existing BWR Feedwater Nozzle B.1.7 existing BWR Penetrations B.1.8 existing BWR Stress Corrosion Cracking B.1.9 existing BWR Vessel ID Attachment Welds B.1.10 existing BWR Vessel Internals B.1.11 existing Compressed Air Monitoring B.1.12 existing Containment Inservice Inspection IWE B.1.13 existing Containment Inservice Inspection IWL B.1.14 existing Containment Leak Rate B.1.15 existing Diesel Fuel Monitoring B.1.16 existing Environmental Qualification (EQ) of Electric Components B.1.17 existing External Surfaces Monitoring B.1.18 existing Fatigue Monitoring B.1.19 existing Fire Protection B.1.20 existing Fire Water System B.1.21 existing Flow-Accelerated Corrosion B.1.22 existing
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-5 Inservice Inspection B.1.23 existing Inservice Inspection-IWF B.1.24 existing Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems B.1.25 existing Internal Surfaces in Miscellaneous Piping and Ducting Components B.1.26 new Masonry Wall B.1.27 existing Non-EQ Cable Connections B.1.28 new Non-EQ Inaccessible Power Cable (400 V to 35 kV)
B.1.29 existing Non-EQ Instrumentation Circuits Test Review B.1.30 new Non-EQ Insulated Cables and Connections B.1.31 new Oil Analysis B.1.32 existing One-Time Inspection B.1.33 new One-Time Inspection - Small-Bore Piping B.1.34 new Periodic Surveillance and Preventive Maintenance B.1.35 existing Protective Coating Monitoring and Maintenance B.1.36 existing Reactor Head Closure Studs B.1.37 existing Reactor Vessel Surveillance B.1.38 existing RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants B.1.39 existing Selective Leaching B.1.40 new Service Water Integrity B.1.41 existing Structures Monitoring B.1.42 existing Water Chemistry Control - BWR B.1.43 existing Table B-1 Aging Management Programs (Continued)
Program Section New or Existing
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-6 B.0.6 CORRELATION WITH NUREG-1801 AGING MANAGEMENT PROGRAMS The correlation between NUREG-1801 programs and GGNS programs is shown below. For the GGNS programs, links to appropriate sections of this appendix are provided.
Water Chemistry Control - Closed Treated Water Systems B.1.44 existing Table B-2 GGNS AMP Correlation with NUREG-1801 Programs NUREG-1801 Number NUREG-1801 Program GGNS Program X.E1 Environmental Qualification (EQ) of Electric Components Environmental Qualification (EQ) of Electric Components [B.1.17]
X.M1 Fatigue Monitoring Fatigue Monitoring [B.1.19]
X.S1 Concrete Containment Tendon Prestress GGNS does not have pre-stressed tendons in the containment structure.
This NUREG-1801 program does not apply.
XI.M1 ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Inservice Inspection - ISI [B.1.23]
XI.M2 Water Chemistry Water Chemistry Control - BWR
[B.1.43]
XI.M3 Reactor Head Closure Stud Bolting Reactor Head Closure Studs [B.1.37]
XI.M4 BWR Vessel ID Attachment Welds BWR Vessel ID Attachment Welds
[B.1.10]
XI.M5 BWR Feedwater Nozzle BWR Feedwater Nozzle [B.1.7]
XI.M6 BWR Control Rod Drive Return Line Nozzle BWR CRD Return Line Nozzle [B.1.6]
Table B-1 Aging Management Programs (Continued)
Program Section New or Existing
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-7 XI.M7 BWR Stress Corrosion Cracking BWR Stress Corrosion Cracking [B.1.9]
XI.M8 BWR Penetrations BWR Penetrations [B.1.8]
XI.M9 BWR Vessel Internals BWR Vessel Internals [B.1.11]
XI.M10 Boric Acid Corrosion GGNS is a BWR. This NUREG-1801 program does not apply.
XI.M11B Cracking of Nickel-Alloy Components and Loss of Material Due to Boric Acid-Induced Corrosion in Reactor Coolant Pressure Boundary Components (PWRs only)
GGNS is a BWR. This NUREG-1801 program does not apply.
XI.M12 Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS)
This NUREG-1801 program is not credited for aging management. Aging effects for CASS components susceptible to thermal aging embrittlement at GGNS are managed by BWR Vessel Internals [B.1.11]
XI.M16A PWR Vessel Internals GGNS is a BWR. This NUREG-1801 program does not apply.
XI.M17 Flow-Accelerated Corrosion Flow-Accelerated Corrosion Program
[B.1.22]
XI.M18 Bolting Integrity Bolting Integrity [B.1.3]
XI.M19 Steam Generators GGNS is a BWR. This NUREG-1801 program does not apply.
XI.M20 Open-Cycle Cooling Water System Service Water Integrity [B.1.41]
XI.M21A Closed Treated Water Systems Water Chemistry Control - Closed Treated Water Systems [B.1.44]
XI.M22 Boraflex Monitoring Boraflex Monitoring [B.1.4]
Table B-2 GGNS AMP Correlation with NUREG-1801 Programs (Continued)
NUREG-1801 Number NUREG-1801 Program GGNS Program
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-8 XI.M23 Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems Inspection of Overhead Heavy Load and Light Load (Related to Refueling)
Handling Systems [B.1.25].
XI.M24 Compressed Air Monitoring Compressed Air Monitoring [B.1.12]
XI.M25 BWR Reactor Water Cleanup System Not credited for aging management.
Refer to relevant discussion in Table 3.3.1, Item 3.3.1-16.
XI.M26 Fire Protection Fire Protection [B.1.20]
XI.M27 Fire Water System Fire Water System [B.1.21]
XI.M29 Aboveground Metallic Tanks Aboveground Metallic Tanks [B.1.2]
XI.M30 Fuel Oil Chemistry Diesel Fuel Monitoring [B.1.16]
XI.M31 Reactor Vessel Surveillance Reactor Vessel Surveillance [B.1.38]
XI.M32 One-Time Inspection One-Time Inspection [B.1.33]
XI.M33 Selective Leaching Selective Leaching [B.1.40]
XI.M35 One-Time Inspection of ASME Code Class 1 Small-Bore Piping One-Time Inspection - Small-Bore Piping [B.1.34]
XI.M36 External Surfaces Monitoring of Mechanical Components External Surfaces Monitoring [B.1.18]
XI.M37 Flux Thimble Tube Inspection GGNS is a BWR. This NUREG-1801 program does not apply.
XI.M38 Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Internal Surfaces in Miscellaneous Piping and Ducting Components
[B.1.26]
XI.M39 Lubricating Oil Analysis Oil Analysis [B.1.32]
Table B-2 GGNS AMP Correlation with NUREG-1801 Programs (Continued)
NUREG-1801 Number NUREG-1801 Program GGNS Program
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-9 XI.M40 Monitoring of Neutron-Absorbing Materials Other than Boraflex There are no neutron-absorbing materials subject to aging management review at GGNS other than Boraflex.
This NUREG-1801 program does not apply.
XI.M41 Buried and Underground Piping and Tanks Buried Piping and Tanks Inspection
[B.1.5]
XI.E1 Insulation Material for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Non-EQ Insulated Cables and Connections [B.1.31]
XI.E2 Insulation Material for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits Non-EQ Instrumentation Circuits Test Review [B.1.30]
XI.E3 Inaccessible Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Non-EQ Inaccessible Power Cables (400V to 35 kV) [B.1.29]
XI.E4 Metal Enclosed Bus GGNS has no metal enclosed bus subject to aging management review.
This NUREG-1801 program does not apply.
XI.E5 Fuse Holders Not credited for aging management.
Refer to relevant discussion in Table 3.6.1, Items 3.6.1-16 and 3.6.1-17.
XI.E6 Electrical Cable Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Non-EQ Cable Connections [B.1.28]
Table B-2 GGNS AMP Correlation with NUREG-1801 Programs (Continued)
NUREG-1801 Number NUREG-1801 Program GGNS Program
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-10 Table B-3 indicates the consistency of GGNS programs with NUREG-1801 programs.
XI.S1 ASME Section XI, Subsection IWE Containment Inservice Inspection IWE [B.1.13]
XI.S2 ASME Section XI, Subsection IWL Containment Inservice Inspection IWL [B.1.14]
XI.S3 ASME Section XI, Subsection IWF Inservice Inspection - IWF [B.1.24]
XI.S4 10 CFR 50, Appendix J Containment Leak Rate [B.1.15]
XI.S5 Masonry Walls Masonry Wall [B.1.27]
XI.S6 Structures Monitoring Structures Monitoring [B.1.42]
XI.S7 RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants [B.1.39]
XI.S8 Protective Coating Monitoring and Maintenance Program Protective Coating Monitoring and Maintenance Program [B.1.36]
Plant-Specific Programs NA Plant-specific program 115 kV Inaccessible Transmission Cable
[B.1.1]
NA Plant-specific program Periodic Surveillance and Preventive Maintenance [B.1.35]
Table B-2 GGNS AMP Correlation with NUREG-1801 Programs (Continued)
NUREG-1801 Number NUREG-1801 Program GGNS Program
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-11 Table B-3 GGNS Program Consistency with NUREG-1801 NUREG-1801 Comparison Program Name Plant Specific Consistent with NUREG-1801 Programs with Enhancements Programs with Exceptions to NUREG-1801 115 kV Inaccessible Transmission Cable X
Aboveground Metallic Tanks X
Bolting Integrity X
X Boraflex Monitoring X
X Buried Piping and Tanks Inspection X
BWR Stress Corrosion Cracking X
BWR Vessel ID Attachment Welds X
BWR Vessel Internals X
X Compressed Air Monitoring X
X Containment Inservice Inspection IWE X
Containment Inservice Inspection IWL X
Containment Leak Rate X
Diesel Fuel Monitoring X
X
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-12 Environmental Qualification (EQ) of Electric Components X
External Surfaces Monitoring X
X Fatigue Monitoring X
X Fire Protection X
X Fire Water System X
X Flow-Accelerated Corrosion X
X Inservice Inspection X
Inservice Inspection - IWF X
X Inspection of Overhead Heavy Load and Light Load (Related to Refueling)
Handling Systems X
X Internal Surfaces in Miscellaneous Piping and Ducting Components X
Masonry Wall X
X Non-EQ Cable Connections X
Non-EQ Inaccessible Power Cables (400V to 35 kV)
X X
Non-EQ Instrumentation Circuits Test Review X
Non-EQ Insulated Cables and Connections X
Oil Analysis X
X Table B-3 GGNS Program Consistency with NUREG-1801 (Continued)
NUREG-1801 Comparison Program Name Plant Specific Consistent with NUREG-1801 Programs with Enhancements Programs with Exceptions to NUREG-1801
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-13 One-Time Inspection X
One-Time Inspection -
Small-Bore Piping X
Periodic Surveillance and Preventive Maintenance X
X Protective Coating Monitoring and Maintenance X
X Reactor Head Closure Studs X
Reactor Vessel Surveillance X
X RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants X
X Selective Leaching X
Service Water Integrity X
Structures Monitoring X
X Water Chemistry Control -
BWR X
Water Chemistry Control -
Closed Treated Water Systems X
X Table B-3 GGNS Program Consistency with NUREG-1801 (Continued)
NUREG-1801 Comparison Program Name Plant Specific Consistent with NUREG-1801 Programs with Enhancements Programs with Exceptions to NUREG-1801
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-14 B.1 AGING MANAGEMENT PROGRAMS AND ACTIVITIES B.1.1 115 KV INACCESSIBLE TRANSMISSION CABLE Program Description There is no corresponding NUREG-1801 program.
The 115 kV Inaccessible Transmission Cable Program is a new condition monitoring program that will manage the effects of aging on the 115 kV inaccessible transmission cables. In this program, inaccessible transmission cables will be tested to provide an indication of the condition of the cable insulation properties. The specific type of test will be a proven test for detecting deterioration of the cable insulation.
This program will be implemented prior to the period of extended operation, and the first cable tests and manhole inspections are to be completed prior to the period of extended operation.
Evaluation
- 1. Scope of Program This program applies to inaccessible underground (e.g., in conduit, duct banks, or direct buried) transmission cables (115 kV) within the scope of license renewal that are exposed to significant moisture. Significant moisture is defined as periodic exposure to moisture that lasts more than a few days (e.g., cable wetting or submergence in water). The following cables are included in the scope of this program:
ESF 12 Feeder - 115 kV Switchyard to Overhead Power Pole: Phases A, B, & C (2 cables per phase)
- 2. Preventive Actions This program will take periodic actions to prevent cables from being exposed to significant moisture by inspecting for water collection in cable manholes and removing water, as needed.
The inspection frequency for manholes will be established and performed based on plant-specific operating experience with cable wetting or submergence in manholes.
Condition-based inspections of manholes not automatically dewatered by a sump pump will be performed based on (a) the potential for high water table conditions and (b) the occurrence of periods of heavy rain. In addition, operation of dewatering devices if applicable will be inspected and operation verified prior to any known or predicted heavy rain or flooding events.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-15 The periodic inspection will occur at least annually. The manhole inspections will include direct observation that cables are not wetted or submerged, that cables/splices and cable support structures are intact, and verification that dewatering/drainage systems (i.e.,
sump pumps) and associated alarms if applicable operate properly. If water is found during inspection (i.e., cable exposed to significant moisture), corrective actions are taken to keep the cable dry and to assess cable degradation. The first inspection for license renewal is completed prior to the period of extended operation.
- 3. Parameters Monitored/Inspected Inspection for water collection in manholes will be performed based on plant-specific operating experience with water accumulation in the manhole, with the inspections to occur at least annually.
In-scope transmission cables (115 kV) that are exposed to significant moisture will be tested to provide an indication of the condition of the conductor insulation. The specific type of test performed will be determined prior to the initial test. The test will be a proven test for detecting deterioration of the insulation system due to wetting or submergence, such as dielectric loss (dissipation factor/power factor), AC voltage withstand, partial discharge, step voltage, time domain reflectometry, insulation resistance and polarization index, line resonance analysis, or other testing that is state-of-the-art at the time the test is performed.
- 4. Detection of Aging Effects Testing will be performed at least once every six years, with the first tests for license renewal occurring before the period of extended operation. For transmission cables exposed to significant moisture, test frequencies may be adjusted based on test results (including trending of degradation where applicable) and operating experience.
The condition of the cable insulation will be assessed with reasonable confidence using test such as dielectric loss (dissipation factor/power factor), AC voltage withstand, partial discharge, step voltage, time domain reflectometry, insulation resistance and polarization index, line resonance analysis, or other testing that is state-of-the-art at the time the tests are performed. The test used to determine the condition of the cable insulation will ensure the cables continue to meet their intended function during the period of extended operation.
- 5. Monitoring and Trending Trending will be used as part of this program based on the ability of trending the test results for the specific test chosen. Since the ability to trend test results will depend on the specific type of test chosen, only results that can be trended will be used to provide additional information on the rate of cable insulation degradation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-16
- 6. Acceptance Criteria The acceptance criteria for each test will be defined by the type of test performed and the specific cable tested. Acceptance criteria for inspections of manholes are defined by the observation that the cables and support structures are not submerged or immersed in standing water at the time of the inspection.
- 7. Corrective Actions An engineering evaluation will be performed when the test acceptance criteria are not met to ensure that the intended functions of the electrical cables can be maintained consistent with the current licensing basis. When an unacceptable condition or situation is identified, a determination is made as to whether the same condition or situation is applicable to other in-scope transmission cables. Corrective actions may include, but are not limited to, installation of permanent drainage systems, installation of sump pumps and alarms, more frequent cable testing or manhole inspections, or replacement of the affected cable.
When an unacceptable condition or situation is identified, the requirements of 10 CFR Part 50, Appendix B, will be used to address corrective actions.
8.
Confirmation Process This element is discussed in Section B.0.3.
9.
Administrative Controls This element is discussed in Section B.0.3.
- 10. Operating Experience The 115 kV Inaccessible Transmission Cable Program is a new program. Industry operating experience was considered in the development of this program. Plant operating experience will be gained as the program is implemented and will be factored into the program via the confirmation and corrective action elements of the GGNS 10 CFR 50 Appendix B quality assurance program.
This inspection program applies to a potential aging effect for which there is no operating experience at GGNS indicating the need for an aging management program. A search of GGNS operating experience with the 115kV inaccessible transmission cables and connections in this program identified no age-related failures and no aging mechanisms not considered in NUREG-1801 have been identified. The GGNS program is similar to the program description in NUREG-1801,Section XI.E3, which in turn is based on industry operating experience that demonstrates that this program is effective for managing the aging effects described herein. As such, operating experience assures that implementation of the 115 kV Inaccessible Transmission Cable Program will manage the
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-17 effects of aging such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Enhancements None Conclusion The 115 kV Inaccessible Transmission Cable Program will be effective for managing aging effects since it will incorporate proven monitoring techniques, acceptance criteria, corrective actions, and administrative controls. The 115 kV Inaccessible Transmission Cable Program provides assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-18 B.1.2 ABOVEGROUND METALLIC TANKS Program Description The Aboveground Metallic Tanks Program is a new program that will manage loss of material for the outer surfaces, including the bottom surfaces, of above-ground metallic tanks constructed on concrete or soil, using periodic visual inspections, measurements of the thickness of the tank bottoms, and preventive measures such as protective coatings and sealants.
This program will be implemented prior to the period of extended operation.
NUREG-1801 Consistency The Aboveground Metallic Tanks Program will be consistent with the program described in NUREG-1801,Section XI.M29, Aboveground Metallic Tanks.
Exceptions to NUREG-1801 None Enhancements None Operating Experience The Aboveground Metallic Tanks Program is a new program. Industry operating experience was considered in the development of this program. Plant operating experience will be gained as the program is implemented and will be factored into the program via the confirmation and corrective action elements of the GGNS 10 CFR 50 Appendix B quality assurance program.
The visual inspection and thickness measurement methods used in this program to detect aging effects are proven industry techniques that have been effectively used at GGNS in other programs. Visual inspections of the condensate storage tank and fire water storage tanks in 2007, 2008, and 2009 found indications of degradation which were resolved prior to any loss of intended function. As such, operating experience assures that implementation of the Aboveground Metallic Tanks Program will manage the effects of aging such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-19 Conclusion The Aboveground Metallic Tanks Program will be effective for managing aging effects since it will incorporate proven monitoring techniques, acceptance criteria, corrective actions, and administrative controls. The Aboveground Metallic Tanks Program provides assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-20 B.1.3 BOLTING INTEGRITY Program Description The Bolting Integrity Program is an existing program that manages loss of preload, cracking, and loss of material for closure bolting for pressure-retaining components using preventive and inspection activities. Applicable industry standards and guidance documents such as NUREG-1339, EPRI NP-5769, and EPRI TR-104213 are used to delineate the program.
NUREG-1801 Consistency The Bolting Integrity Program, with enhancements, is consistent with the program described in NUREG-1801,Section XI.M18, Bolting Integrity.
Exceptions to NUREG-1801 None Enhancements The following enhancements will be implemented prior to the period of extended operation.
Elements Affected Enhancement
- 2. Preventive Actions The Bolting Integrity Program will be enhanced to clarify the prohibition on use of lubricants containing MoS2 for bolting and to specify that proper gasket compression will be visually verified following assembly. The scope of this enhancement will include applicable GGNS site procedures.
- 4. Detection of Aging Effects The Bolting Integrity Program will be enhanced to include consideration of the guidance applicable for pressure boundary bolting in NUREG-1339, EPRI NP-5769, and EPRI TR-104213.
- 4. Detection of Aging Effects The Bolting Integrity Program will be enhanced to include volumetric examination per ASME Code Section XI, Table IWB-2500-1, Examination Category B-G-1, for high-strength closure bolting regardless of code classification. High strength closure bolting is that with an actual yield strength greater than or equal to 150 ksi.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-21 Operating Experience Class I bolting degradation has been observed in control rod drive (CRD) cap screws, as initially reported by General Electric. Extensive evaluations were performed to understand the degradation mechanism and its affect on cap screw integrity. Evaluation and supporting analysis concluded in 1998 that CRD cap screw degradation is caused by corrosion. Through analysis and metallurgical evaluation, the condition has been proven to be acceptable for long term operation.
Corrosion was observed in 2002 on the studs for a flange in standby service water (SSW) system piping, located in the "B" SSW basin under approximately 5 feet of water. The flange studs, nuts, and portions of the flange were observed to be covered with an iron colored deposit. When the deposit was removed from the studs and nuts on this flange, most of the protective coating was found to be deteriorated and there was noticeable metal loss from the studs. There were no signs of system leakage at this location nor were there any signs of previous leakage. The corroded flange bolts were replaced.
During an inspection of reactor recirculation pump heat exchanger studs in 2005, the presence of some light pitting and thread loss due to corrosion was reported. The studs were replaced. An inspection of the removed studs by engineering and additional mechanical maintenance personnel concluded that no loss of intended function had occurred.
While performing examinations of core spray sparger bolts in 2005, a tack weld on one bolt was found to be cracked. The tack weld is provided to prevent de-tensioning of the bolt, serving the same function as a lock wire. The remaining configuration, including another tack weld on this bolt, was found to provide an adequate locking mechanism.
A slight water leak was detected in 2006 at a bolt on the pump seal retaining ring for the high pressure core spray pump. The bolt was cleaned and retorqued.
The Bolting Integrity Program has been effective in identification of conditions and program deficiencies. Appropriate corrective actions have been implemented to correct program deficiencies and to ensure future integrity of the bolted connections. This provides assurance that the program will remain effective for managing loss of material. The history of identification of degradation and initiation of corrective action prior to loss of intended function provide assurance that the program is effective for managing aging effects for passive components.
- 7. Corrective Actions The Bolting Integrity Program will be enhanced to include guidance from EPRI NP-5769 and EPRI TR-104213 for replacement of bolting.
Elements Affected Enhancement
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-22 The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The Bolting Integrity Program has been effective at managing aging effects. The Bolting Integrity Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-23 B.1.4 BORAFLEX MONITORING Program Description The Boraflex Monitoring Program is an existing program that manages the change in material properties (neutron-absorbing capacity) in the Boraflex material affixed to spent fuel racks using silica sampling, areal density testing, and other monitoring activities. Inspection frequency and acceptance criteria are based on the GGNS response to NRC Generic Letter 96-04 and the GGNS technical specifications.
NUREG-1801 Consistency The Boraflex Monitoring Program, with enhancements, is consistent with the program described in NUREG-1801,Section XI.M22, Boraflex Monitoring.
Exceptions to NUREG-1801 None Enhancements The following enhancement will be implemented prior to the period of extended operation.
Elements Affected Enhancement 3.
Parameters Monitored or Inspected 4.
Detection of Aging Effects 5.
Monitoring and Trending GGNS will perform periodic surveillances of the Boraflex neutron absorbing material on at least a five-year frequency using Boron-10 Areal Density Gage for Evaluating Racks (BADGER) testing.
RACKLIFE analysis will continue to be performed each cycle. This analysis will include a comparison of the RACKLIFE predicted silica to the plant measured silica. This comparison will determine if adjustments to the RACKLIFE loss coefficient are merited. The analysis will include projections to the next planned RACKLIFE analysis date to ensure current Region I storage locations will not need to be reclassified as Region II storage locations in the analysis interval.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-24 Operating Experience The results of spent fuel storage pool neutron transmission testing in 1999 revealed boraflex gaps in the test area exceeding those assumed in the criticality safety analysis. Engineering disposition of these results was used to prohibit storage of any fuel in the spent fuel pool boraflex test area.
In-situ measurement of the boron-10 areal density of the neutron absorber material in 2007 identified no immediate actions or concerns for the spent fuel pool rack criticality safety analysis, although future limitations on the use of the spent fuel pool storage areas were indicated.
The Boraflex Monitoring Program has been effective in identification of conditions and program deficiencies. Appropriate corrective actions have been implemented. This provides assurance that the program will remain effective for managing loss of material. The history of identification of degradation and initiation of corrective action prior to loss of intended function provide assurance that the program is effective for managing aging effects for passive components.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The Boraflex Monitoring Program has been effective at managing aging effects. The Boraflex Monitoring Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-25 B.1.5 BURIED PIPING AND TANKS INSPECTION Program Description The Buried Piping and Tanks Inspection Program is a new program that manages loss of material for the external surfaces of buried and underground piping and tanks composed of any material through preventive, mitigative, and inspection activities.
This program will be implemented prior to the period of extended operation.
NUREG-1801 Consistency The Buried Piping and Tanks Inspection Program will be consistent with the program described in NUREG-1801,Section XI.M41, Buried and Underground Piping and Tanks.
Exceptions to NUREG-1801 None Enhancements None Operating Experience The Buried Piping and Tanks Inspection Program is a new program. Industry operating experience will be considered when implementing this program. Plant operating experience for this program will be gained as it is implemented during the period of extended operations and will be factored into the program via the confirmation and corrective action elements of the GGNS 10 CFR 50 Appendix B quality assurance program.
Prior to installation of a new cathodic protection system at GGNS in December of 2009, a native area potential earth current inspection was performed throughout the protected area to establish baseline readings. No degraded conditions were identified, although two areas were noted to have a higher potential for coating degradation.
The GGNS program will be based on the program description in NUREG-1801, which in turn is based on industry operating experience that demonstrates that this program is effective for managing the aging effects described herein. As such, operating experience assures that implementation of the Buried Piping and Tanks Inspection Program will manage the effects of aging such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-26 Conclusion The Buried Piping and Tanks Inspection Program will be effective for managing aging effects since it will incorporate proven monitoring techniques, acceptance criteria, corrective actions, and administrative controls. The Buried Piping and Tanks Inspection Program provides assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-27 B.1.6 BWR CRD RETURN LINE NOZZLE Program Description The BWR Control Rod Drive (CRD) Return Line Nozzle Program is an existing program that manages cracking of the control rod drive return line nozzle using preventive, mitigative, and inservice inspection activities, in accordance with GGNS commitments to Generic Letter 80-095 to implement the recommendations in NUREG-0619.
NUREG-1801 Consistency The BWR CRD Return Line Nozzle Program is consistent with the program described in NUREG-1801,Section XI.M6, BWR Control Rod Drive Return Line Nozzle.
Exceptions to NUREG-1801 None Enhancements None Operating Experience During RF13 in 2004, the N10 nozzle was inspected using ultrasonic examination on the CRD return line nozzle end cap to carbon steel safe end dissimilar metal weld. The results revealed no crack indication. Absence of aging effects indicates that the preventive actions of the program have been effective.
The BWR CRD Return Line Nozzle Program detects aging effects using nondestructive examination visual, surface and volumetric techniques to detect and characterize flaws. These techniques are widely used and have been demonstrated effective at detecting aging effects during inspections performed to meet ASME Section XI Code requirements. In addition, staff-approved BWRVIP documents are based on industry-wide experience at BWR plants. The application of these proven methods provides assurance that the effects of aging will be managed such that the CRD return line nozzle components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-28 Conclusion The BWR CRD Return Line Nozzle Program has been effective at managing aging effects. The BWR CRD Return Line Nozzle Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-29 B.1.7 BWR FEEDWATER NOZZLE Program Description The BWR Feedwater Nozzle Program is an existing program that manages cracking of the BWR feedwater nozzles using inspection activities. This program augments the examinations specified in the ASME Code,Section XI, with the recommendation of General Electric (GE) NE-523-A71-0594 to perform periodic inspection of critical regions of the BWR feedwater nozzles.
NUREG-1801 Consistency The BWR Feedwater Nozzle Program is consistent with the program described in NUREG-1801,Section XI.M5, BWR Feedwater Nozzle.
Exceptions to NUREG-1801 None Enhancements None Operating Experience During RF07 in 1995, volumetric exams were performed on all six feedwater nozzle blend radii as well as on the nozzle-to-safe end welds. Non-relevant geometric reflection indication, inner diameter root geometry, and inside surface geometry were recorded on some of the blend radii and nozzle-to-safe end welds. No crack-like indications were detected. Final analysis of the ultrasonic examination data was performed and found to be acceptable in accordance with ASME Section XI.
During RF09 in 1998, volumetric exams were performed on all six feedwater nozzle blend radii as well as on two nozzle-to-safe end welds. No indications that required evaluation were recorded.
During RF12 in 2002, volumetric exams were performed on all six feedwater nozzle blend radii as well as on two nozzle-to-safe end welds. No indications that required evaluation were recorded.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-30 The BWR Feedwater Nozzle Program detects aging effects using nondestructive examination (NDE) visual, surface and volumetric techniques to detect and characterize flaws. These techniques are widely used and have been demonstrated effective at detecting aging effects during inspections performed to meet ASME Section XI Code requirements. The application of these proven methods provides assurance that the effects of aging will be managed such that the BWR feedwater nozzle components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The BWR Feedwater Nozzle Program has been effective at managing aging effects. The BWR Feedwater Nozzle Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-31 B.1.8 BWR PENETRATIONS Program Description The BWR Penetrations Program is an existing program that manages cracking of BWR vessel penetrations using inspection and flaw evaluation activities. Applicable industry standards and staff-approved BWRVIP documents are used to delineate the program.
NUREG-1801 Consistency The BWR Penetrations Program is consistent with the program described in NUREG-1801,Section XI.M8, BWR Penetrations.
Exceptions to NUREG-1801 None Enhancements None Operating Experience Visual inspections of the instrument penetrations and extensions were performed during 1996 to 2006 with no indications of degradation noted. Absence of aging effects indicates that the preventive actions of the program have been effective.
The BWR Penetrations Program detects aging effects using nondestructive examination techniques to detect and characterize flaws. These techniques are widely used and have been demonstrated effective at detecting aging effects during inspections performed to meet ASME Section XI Code requirements. In addition, the BWRVIP programs are based on industry-wide experience at BWR plants. The application of these proven methods provides assurance that the effects of aging will be managed such that the BWR Penetrations Program components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The BWR Penetrations Program has been effective at managing aging effects. The BWR Penetrations Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-32 B.1.9 BWR STRESS CORROSION CRACKING Program Description The BWR Stress Corrosion Cracking Program is an existing program that manages cracking of the reactor coolant pressure boundary using preventive measures, inspection, and flaw evaluation. Staff-approved BWRVIP documents and the GGNS response to NUREG-0313 Revision 2 and NRC Generic Letter 88-01 and its Supplement 1 are used to delineate the program.
NUREG-1801 Consistency The BWR Stress Corrosion Cracking Program is consistent with the program described in NUREG-1801,Section XI.M7, BWR Stress Corrosion Cracking.
Exceptions to NUREG-1801 None Enhancements None Operating Experience A review of Owner's Activity Reports for 2004 through 2009 showed no indications of cracking from inspections performed under this program. Absence of aging effects indicates that the preventive actions of the program have been effective.
The BWR Stress Corrosion Cracking Program detects aging effects using NDE visual, surface and volumetric techniques to detect and characterize flaws. These techniques are widely used and have been demonstrated effective at detecting aging effects during inspections performed to meet ASME Section XI Code requirements. The application of these proven methods provides assurance that the effects of aging will be managed such that the BWR Stress Corrosion Cracking Program components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-33 Conclusion The BWR Stress Corrosion Cracking Program has been effective at managing aging effects. The BWR Stress Corrosion Cracking Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-34 B.1.10 BWR VESSEL ID ATTACHMENT WELDS Program Description The BWR Vessel ID Attachment Welds Program is an existing program that manages cracking in structural welds for BWR reactor vessel internal integral attachments using inspection and flaw evaluation. Applicable industry standards and staff-approved BWRVIP documents are used to delineate the program.
NUREG-1801 Consistency The BWR Vessel ID Attachment Welds Program is consistent with the program described in NUREG-1801,Section XI.M4, BWR Vessel ID Attachment Welds.
Exceptions to NUREG-1801 None Enhancements None Operating Experience Inspection of reactor vessel internal attachment welds are governed by BWRVIP-48-A.
During RF14 in 2005, twelve jet pump wedges and six other jet pump weld locations were visually examined with no indications noted. Enhanced visual inspection was performed on twenty-four piping welds and six piping brackets with no indications noted. Visual examinations were performed on sixteen core spray brackets, and a tack weld on a bolt was identified as cracked. This condition was found to be acceptable, and the inspection interval for the bolt was revised to each refueling outage. An examination of steam dryer support and hold-down attachment welds was performed with no indications noted.
During RF15 in 2007, enhanced visual inspection was performed on four jet pump wedges and five jet pump riser brace attachment welds, and volumetric examination was performed on nineteen jet pump beams. Enhanced visual inspection of fourteen piping welds was performed.
No indications were noted during these inspections.
Visual inspections of upper guide rod bracket attachment welds and surveillance sample holder bracket attachment welds have identified no indications.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-35 The BWR Vessel ID Attachment Welds Program detects aging effects using NDE visual, surface and volumetric techniques to detect and characterize flaws. These techniques are widely used and have been demonstrated effective at detecting aging effects during inspections performed to meet ASME Section XI Code requirements. In addition, BWRVIP-48-A is based on industry-wide experience at BWR plants. The application of these proven methods provides assurance that the effects of aging will be managed such that the BWR Vessel ID Attachment Welds Program components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The BWR Vessel ID Attachment Welds Program has been effective at managing aging effects.
The BWR Vessel ID Attachment Welds Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-36 B.1.11 BWR VESSEL INTERNALS Program Description The BWR Vessel Internals Program is an existing program that manages cracking, loss of material, and reduction of fracture toughness for BWR vessel internal components using inspection and flaw evaluation. This program also provides (1) determination of the susceptibility of cast austenitic stainless steel components, (2) accounting for the synergistic effect of thermal aging and neutron irradiation, and (3) implementation of a supplemental examination program, as necessary. Applicable industry standards and staff-approved BWRVIP documents are used to delineate the program.
NUREG-1801 Consistency The BWR Vessel Internals Program, with enhancements, is consistent with the program described in NUREG-1801,Section XI.M9, BWR Vessel Internals.
Exceptions to NUREG-1801 None Enhancements The following enhancement will be implemented prior to the period of extended operation.
Elements Affected Enhancements
- 4. Detection of Aging Effects The GGNS program will be enhanced as follows.
(a) The susceptibility to neutron or thermal embrittlement for reactor vessel internal components composed of CASS, X-750 alloy, precipitation-hardened (PH) martensitic stainless steel (e.g., 15-5 and 17-4 PH steel),
and martensitic stainless steel (e.g., 403, 410, 431 steel) will be evaluated.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-37 Operating Experience During RF14 in 2005, volumetric examinations of the core shroud revealed one indication on the lower side of weld H4 with characteristics associated with stress corrosion cracking. This was evaluated as acceptable and included into the core shroud re-inspection intervals. Examination results during 1995 through 2005 allowed the Grand Gulf shroud to remain classified as a category B shroud. Stress calculations and required distributed ligament evaluations were performed on those welds where greater than 50% of the circumference could not be examined.
Enhanced visual inspections performed on shroud support and access hole covers in 2008 during RF16 revealed no indications. No adverse conditions have been noted.
During RF16 in 2008, visual inspection of the core spray sparger revealed no indications.
Cracked tack welds at two cap screws were identified in1998 and 2005 with no further cracking noted in 2008. The locations identified with cracked tack welds were found to be acceptable.
(cont.)
(b) Portions of the susceptible components determined to be limiting from the standpoint of thermal aging susceptibility, neutron fluence, and cracking susceptibility (i.e.,
applied stress, operating temperature, and environmental conditions) will be inspected, using an inspection technique capable of detecting the critical flaw size with adequate margin. The critical flaw size will be determined based on the service loading condition and service-degraded material properties. The initial inspection will be performed either prior to or within 5 years after entering the period of extended operation. If cracking is detected after the initial inspection, the frequency of re-inspection will be justified based on fracture toughness properties appropriate for the condition of the component. The sample size will be 100% of the accessible component population, excluding components that may be in compression during normal operations.
Elements Affected Enhancements
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-38 During RF09 in 1998, GGNS began implementation of BWRVIP-42 requirements for baseline inspections of low pressure coolant injection (LPCI) coupling assemblies. These were completed with no indications or cracks identified. Further baseline inspections were completed during RF10 in 1999 and again during RF12 in 2002 with no indications or cracks identified. Inspections of LPCI coupling assemblies continued during RF13 and RF14 in 2004 and 2005 with no indications or cracks identified.
Enhanced visual inspections of control rod internal housings, control rod guide tubes, and stub tubes during RF16 in 2008 were performed with no indications noted.
Multiple inspections of incore housing, guide tubes and dry tubes have been performed, with no indications. This includes inspection of the top two feet of the SRM/IRM dry tubes. The initial dry tube inspection was accomplished during RF02 in 1987. During RF16 in 2008, twenty-four LPRM dry tubes were inspected with no indications noted.
During RF16 in 2008, jet pump assemblies were inspected with no indications of cracking, but slight wear was noted on the wedge rod at jet pumps 01, 02, 05, 06, 07, and 09. This wear was evaluated as acceptable with future inspections required. The examinations conducted to date on the jet pump assembly welds are adequate to assure the intended function of the jet pump assemblies. All jet pump beams were replaced with improved heat treated beams in response to USNRC IEB 80-07. To date no indications have been noted on the jet pump beams.
Steam dryer examinations have been performed since RF01 in 1986. During RF03 in 1989, damage was located on the lower guide and a 1/8" crack was identified on vertical bank weld V5.
Reinspection of the identified crack during RF04 and RF05 in 1990 and 1992 showed no change.
Several examinations were performed on various dryer components at each refueling outage beginning with RF06 in 1993 through RF14 in 2005. During RF15 in 2007, the BWRVIP-139 baseline examinations were completed on all external welds. Broken tack welds on the lifting lugs and indications on the upper support ring were identified. During RF16 in 2008, the indications noted during RF15 were inspected again. An additional indication was identified on the tack weld at 220° lifting lug and an additional indication on the upper support ring. These indications were found to be acceptable for continued operation.
The BWR Vessel Internals Program detects aging effects using NDE visual, surface and volumetric techniques to detect and characterize flaws. These techniques are widely used and have been demonstrated effective at detecting aging effects during inspections performed to meet ASME Section XI Code requirements. In addition, the various BWRVIPs applied in this program are based on industry-wide experience at BWR plants. The application of these proven methods provides assurance that the effects of aging will be managed such that the BWR Vessel Internals Program components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-39 Conclusion The BWR Vessel Internals Program has been effective at managing aging effects. The BWR Vessel Internals Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-40 B.1.12 COMPRESSED AIR MONITORING Program Description The Compressed Air Monitoring Program is an existing program that manages loss of material in compressed air systems by monitoring air samples for moisture and contaminants and by inspecting internal surfaces within compressed air systems. Inspection frequency and acceptance criteria are based on the GGNS response to NRC Generic Letter 88-14 along with applicable industry standards and guidance documents.
NUREG-1801 Consistency The Compressed Air Monitoring Program, with enhancements, is consistent with the program described in NUREG-1801,Section XI.M24, Compressed Air Monitoring.
Exceptions to NUREG-1801 None Enhancements The following enhancements will be implemented prior to the period of extended operation.
Elements Affected Enhancements
- 2. Preventive Actions The GGNS Compressed Air Monitoring Program will be enhanced to apply a consideration of the guidance of ASME OM-S/G-1998, Part 17; American National Standards Institute (ANSI)/ISA-S7.0.01-1996; EPRI NP-7079; and EPRI TR-108147 to the limits specified for air system contaminants.
- 3. Parameters Monitored or Inspected
- 5. Monitoring and Trending The GGNS Compressed Air Monitoring Program will be enhanced to include periodic and opportunistic inspections of accessible internal surfaces of piping and components in the following compressed air systems.
- Division 1 Diesel Generator Starting Air (D1DGSA)
- Division 2 Diesel Generator Starting Air (D2DGSA)
- Division 3 Diesel Generator Starting Air (D3DGSA),
also known as the HPCS Diesel Generator
- Instrument Air (IA) - system P53
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-41 Operating Experience Evidence of rust was found in 2003 during internal inspection of the standby diesel generator starting air tanks. The rust did not impact the wall thickness of the tanks, and no loose products were observed. Actions were taken to remove the rust. Identification of aging effects and corrective actions prior to loss of intended function indicates that the inspection activities of the program are effective.
In 2009, a concern was identified with high dew points for diesel generator starting air systems (Division 1, 2, and 3) for the previous two years. Corrective actions included creating new repetitive tasks for maintenance on the air dryers and revising the procedure for desiccant replacement. The response to a high dew point reading now requires a check of the dryer tower crossover valves and a satisfactory dryer retest after replacement of the desiccant. This procedure ensures that the preventive actions for this program remain effective in managing aging effects.
Instrument air samples collected in 2010 exceeded the procedural limit for particulate size. The cause was attributed to filters on temporary air compressors used during RF17. A system modification was implemented to remove the temporary air compressors from the system.
Identification of program deficiencies and subsequent corrective actions provide assurance that the Compressed Air Monitoring Program will remain effective for managing loss of material of components. The application of these proven methods provides assurance that the effects of aging will be managed such that components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The Compressed Air Monitoring Program has been effective at managing aging effects. The Compressed Air Monitoring Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-42 B.1.13 CONTAINMENT INSERVICE INSPECTION - IWE Program Description The Containment Inservice Inspection - IWE Program is an existing program that performs a general visual examination to assess the condition of the containment steel liner and to detect evidence of degradation that may affect structural integrity or leak tightness. This examination satisfies the requirements of the ASME Boiler and Pressure Vessel Code (to include the 1998 edition with 1999 and 2000 addenda, 2001 edition with 2003 addenda, and the 2004 Code Edition),Section XI, Subsection IWE Examination Category E-A.
The program is augmented by existing plant procedures to ensure that the selection of bolting material installation torque or tension and the use of lubricants and sealants is appropriate for the intended purpose. These procedures reference guidance contained in EPRI TR-104213, NUREG-1339 and EPRI NP-5769 to ensure proper specification of bolting material, lubricant, and installation torque.
NUREG-1801 Consistency The Containment Inservice Inspection - IWE Program is consistent with the program described in NUREG-1801,Section XI.S1, ASME Section XI, Subsection IWE.
Exceptions to NUREG-1801 None Enhancements None.
Operating Experience During a general visual inspection of the containment liner plate in 2003, a gouge was identified.
An engineering review concluded that the liner plate was still capable of performing its intended function to provide a leak tight barrier after a design basis accident. The gouge was repaired during RF13 in 2004. Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing aging effects.
Visual examination of the containment liner in 2007 revealed flaking, blistering, peeling, and chipping conditions that were acceptable as is. Identification of degradation prior to loss of intended function provides evidence that the program is effective for managing aging effects.
Examination results for the containment liner, fuel transfer tube, and other components during RF17 in 2010 were acceptable.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-43 The Containment Inservice Inspection - IWE Program detects aging effects using nondestructive examination visual surface techniques to detect and characterize flaws. These techniques are widely used and have been demonstrated effective at detecting aging effects during inspections performed to meet ASME Section XI Code requirements. Identification of program deficiencies and subsequent corrective actions provide assurance that the program will remain effective for managing loss of material of components.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The Containment Inservice Inspection - IWE Program has been effective at managing aging effects. The Containment Inservice Inspection - IWE Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-44 B.1.14 CONTAINMENT INSERVICE INSPECTION - IWL Program Description The Containment Inservice Inspection - IWL Program is an existing program that performs a general visual examination to assess the overall condition of the containment concrete and to detect evidence of degradation that may affect structural integrity or leak tightness. These examinations are used to meet the examination requirements of the ASME Boiler and Pressure Vessel Code (1998 Edition with the 2000 Addenda, 2001 Edition through the 2003 Addenda, and 2004 Edition)Section XI, Subsection IWL Examination Category L-A, Item Numbers L1.11, L1.12, and L2.30. In accordance with GGNS-specific relief requests, these examinations are also used as an alternative to the examinations specified in the 1992 edition with 1992 addenda for IWL Examination Category L-A.
NUREG-1801 Consistency The Containment Inservice Inspection - IWL Program is consistent with the program described in NUREG-1801 XI.S2, ASME Section XI, Subsection IWL.
Exceptions to NUREG-1801 None Enhancements None Operating Experience Inspections of concrete are performed consistent with the schedule outlined in the ISI program.
A review of Owner's Activity Reports for 2004 through 2009 showed no adverse indications from these inspections.
The Containment Inservice Inspection - IWL Program detects aging effects using nondestructive examination visual surface techniques to detect and characterize flaws. These techniques are widely used and have been demonstrated effective at detecting aging effects during inspections performed to meet ASME Section XI Code requirements. Identification of program deficiencies, and subsequent corrective actions, provide assurance that the program will remain effective for managing loss of material of components. The application of these proven methods provides assurance that the effects of aging will be managed such that components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-45 Conclusion The Containment Inservice Inspection - IWL Program has been effective at managing aging effects. The Containment Inservice Inspection - IWL Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-46 B.1.15 CONTAINMENT LEAK RATE Program Description The Containment Leak Rate Program is an existing program that provides for detection of loss of material, cracking, and loss of function in various systems penetrating containment. The program also provides for detection of age-related degradation in material properties of gaskets, o-rings, and packing materials for the primary containment pressure boundary access points.
Containment leakage rate tests (LRT) are performed to assure that leakage through the containment and systems and components penetrating primary containment does not exceed allowable leakage limits specified in the plant technical specifications. An integrated leak rate test (ILRT) is performed during a period of reactor shutdown at the frequency specified in 10 CFR Part 50, Appendix J, Option B. Performance of the integrated leak rate test per 10 CFR Part 50, Appendix J demonstrates the leak-tightness and structural integrity of the containment. Local leak rate tests (LLRT) are performed on isolation valves and containment access penetrations at frequencies that comply with the requirements of 10 CFR Part 50, Appendix J, Option B.
NUREG-1801 Consistency The Containment Leak Rate Program is consistent with the program described in NUREG-1801,Section XI.S4, 10 CFR Part 50, Appendix J.
Exceptions to NUREG-1801 None Enhancements None Operating Experience Local leak rate testing (LLRT) during RF10 through RF15 (1999 through 2007) met test acceptance criteria, yet some components did not meet administrative limits. Some of these components were repaired and retested as acceptable, while others were evaluated and deferred. In each of these cases, containment leakage was within overall allowed limits. This indicates that the program is effective at identifying and managing aging effects on primary containment components.
In 2006, the containment isolation valves for penetration #35 failed their LLRT with 3487 sccm leakage against an allowable limit of 3400 sccm. This identified leakage was found acceptable through engineering evaluation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-47 In 2007, the containment equipment hatch failed an LLRT being performed to support hatch reinstallation. Since this was not an as-found test, containment integrity was not considered to be lost. The hatch was reinstalled and the subsequent LLRT was successful.
In 2008, the LLRT on the containment isolation valves for penetration #49 for the filter/
demineralizer system indicated leakage approximately 12 times the administrative limit. A flush of the system was completed, and the valves were then re-tested satisfactorily. A new procedure was established to ensure that a system flush will be completed satisfactorily after future resin transfers.
During the containment integrated leak rate testing in 2008, test data met applicable test acceptance criteria and confirmed the structural integrity of the containment.
A program self-assessment in 2009 revealed a decline in performance due to organizational weaknesses. Follow-up actions from the 2009 self-assessment included improved data analyses and performance monitoring. Reviews against established program standards provide assurance that the program will remain effective for managing loss of material of components.
Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing aging effects for components.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The Containment Leak Rate Program has been effective at managing aging effects. The Containment Leak Rate Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-48 B.1.16 DIESEL FUEL MONITORING Program Description The Diesel Fuel Monitoring Program is an existing program that manages loss of material and fouling in piping and components exposed to an environment of diesel fuel oil by verifying the quality of fuel oil and controlling fuel oil contamination as well as periodic draining, cleaning, and inspection of tanks. Applicable industry standards and guidance documents are used to delineate the program.
The One-Time Inspection Program describes inspections planned to verify that the Diesel Fuel Monitoring Program has been effective at managing aging effects.
NUREG-1801 Consistency The Diesel Fuel Monitoring Program, with enhancements, is consistent with the program described in NUREG-1801,Section XI.M30, Fuel Oil Chemistry.
Exceptions to NUREG-1801 None.
Enhancements The following enhancements will be implemented prior to the period of extended operation.
Elements Affected Enhancements
- 4. Detection of Aging Effects The Diesel Fuel Monitoring Program will be enhanced to include a ten-year periodic cleaning and internal inspection of the fire water pump diesel fuel oil tanks (SP64A002A/B), the diesel fuel oil day tanks for Divisions I, II, III, and the diesel fuel oil drip tanks for Divisions I, II. These cleanings and internal inspections will be performed at least once during the ten-year period prior to the period of extended operation and at succeeding ten-year intervals. If visual inspection is not possible, a volumetric inspection will be performed.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-49 Operating Experience Standby diesel fuel oil storage tank (Division I/II) inspections in 2004 indicated tank internal surfaces were satisfactory. In 2005, sampling of the fire water diesel fuel oil storage tanks indicated no signs of water or foreign material in the tanks. Absence of aging effects indicates that the preventive actions of the program have been effective.
In 2003, inspection of the high pressure core spray (HPCS Division III) diesel fuel oil storage tank indicated small blemishes in the coating. This was corrected prior to return to service.
In 2005, biotrend samples from the fire water diesel fuel oil day tanks measured biological growth on day 3 of a 6-day incubation period. Biocide addition was performed. No further actions were required in this incident. In 2007, biotrend analysis of a sample from the Division I diesel fuel oil day tank showed microbial growth. Biocide addition was performed. Routine sampling and the decisions to add biocide shows the effectiveness of the program.
Results from sampling of the fire water diesel fuel oil storage tanks during 2008 to 2010 for water and sediment were acceptable.
Results from sampling of standby diesel fuel oil storage tanks (Division I/II) and the high pressure core spray diesel fuel oil storage tank (HPCS Division III) during 2006 to 2010 for water and sediment were acceptable.
Identification of conditions and subsequent corrective actions provide assurance that the program will remain effective for managing loss of material of components.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
- 4. Detection of Aging Effects The Diesel Fuel Monitoring Program will be enhanced to include a volumetric examination of affected areas of the diesel fuel tanks if evidence of degradation is observed during visual inspection. The scope of this enhancement includes the diesel fuel oil day tanks (Divisions I, II, III), the diesel fuel oil storage tanks (Divisions I, II, III), the diesel fuel oil drip tanks (Divisions I, II),
and the diesel fire pump fuel oil storage tanks, and is applicable to the inspections performed during the ten-year period prior to the period of extended operation and at succeeding ten-year intervals.
Elements Affected Enhancements
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-50 Conclusion The Diesel Fuel Monitoring Program has been effective at managing aging effects. The Diesel Fuel Monitoring Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-51 B.1.17 ENVIRONMENTAL QUALIFICATION (EQ) OF ELECTRIC COMPONENTS Program Description The Environmental Qualification (EQ) of Electric Components Program is an existing program.
The Nuclear Regulatory Commission (NRC) has established nuclear station environmental qualification (EQ) requirements in 10 CFR Part 50, Appendix A, Criterion 4, and 10 CFR 50.49.
10 CFR 50.49 specifically requires that an EQ program be established to demonstrate that certain electrical components located in harsh plant environments (that is, those areas of the plant that could be subject to the harsh environmental effects of a loss of coolant accident
[LOCA], high energy line breaks [HELBs] or high radiation) are qualified to perform their safety function in those harsh environments. 10 CFR 50.49 requires that the effects of significant aging mechanisms be addressed as part of environmental qualification.
The GGNS EQ program manages the effects of thermal, radiation, and cyclic aging through the use of aging evaluations based on 10 CFR 50.49(f) qualification methods. As required by 10 CFR 50.49, EQ components are refurbished, replaced, or their qualification is extended prior to reaching the aging limits established in the evaluation. Reanalysis of an aging evaluation addresses attributes of analytical methods, data collection and reduction methods, underlying assumptions, acceptance criteria, and corrective actions. Some aging evaluations for EQ components are time-limited aging analyses (TLAAs) for license renewal.
EQ Component Reanalysis Attributes The reanalysis of an aging evaluation is normally performed to extend the qualification by reducing excess conservatism incorporated in the prior evaluation. Reanalysis of an aging evaluation to extend the qualification of a component is performed on a routine basis pursuant to 10 CFR 50.49(e) as part of an EQ program. While a component life limiting condition may be due to thermal, radiation, or cyclical aging, the vast majority of component aging limits are based on thermal conditions. Conservatism may exist in aging evaluation parameters, such as the assumed ambient temperature of the component, an unrealistically low activation energy, or in the application of a component (de-energized versus energized). The reanalysis of an aging evaluation is documented according to the station's quality assurance program requirements that require the verification of assumptions and conclusions. As already noted, important attributes of a reanalysis include analytical methods, data collection and reduction methods, underlying assumptions, acceptance criteria, and corrective actions (if acceptance criteria are not met).
These attributes are discussed below.
Analytical Methods: The analytical models used in the reanalysis of an aging evaluation are the same as those applied during the prior evaluation. The Arrhenius methodology is an acceptable thermal model for performing a thermal aging evaluation. The analytical method used for a radiation aging evaluation is to demonstrate qualification for the total integrated dose (that is, normal radiation dose for the projected installed life plus accident radiation dose). For license renewal, one acceptable method of establishing the 60-year normal radiation dose is to multiply
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-52 the 40-year normal radiation dose by 1.5 (that is, 60 years/40 years). The result is added to the accident radiation dose to obtain the total integrated dose for the component. For cyclical aging, a similar approach may be used. Other models may be justified on a case-by-case basis.
Data Collection and Reduction Methods: Reducing excess conservatism in the component service conditions (for example, temperature, radiation, cycles) used in the prior aging evaluation is the chief method used for a reanalysis. Temperature data used in an aging evaluation is to be conservative and based on plant design temperatures or on actual plant temperature data.
When used, plant temperature data can be obtained in several ways, including monitors used for Technical Specification compliance, other installed monitors, measurements made by plant operators during rounds, and temperature sensors on large motors (while the motor is not running). A representative number of temperature measurements are conservatively evaluated to establish the temperatures used in an aging evaluation. Plant temperature data may be used in an aging evaluation in different ways, such as (a) directly applying the plant temperature data in the evaluation, or (b) using the plant temperature data to demonstrate conservatism when using plant design temperatures for an evaluation. Any changes to material activation energy values as part of a reanalysis are to be justified on a plant-specific basis. Similar methods of reducing excess conservatism in the component service conditions used in prior aging evaluations can be used for radiation and cyclical aging.
Underlying Assumptions: EQ component aging evaluations contain sufficient conservatism to account for most environmental changes occurring due to plant modifications and events. When unexpected adverse conditions are identified during operational or maintenance activities that affect the normal operating environment of a qualified component, the affected EQ component is evaluated and appropriate corrective actions are taken that may include changes to the qualification bases and conclusions.
Acceptance Criteria and Corrective Actions: The reanalysis of an aging evaluation could extend the qualification of the component. If the qualification cannot be extended by reanalysis, the component is to be refurbished, replaced, or requalified prior to exceeding the period for which the current qualification remains valid. A reanalysis is to be performed in a timely manner (that is, sufficient time is available to refurbish, replace, or requalify the component if the reanalysis is unsuccessful).
NUREG-1801 Consistency The Environmental Qualification (EQ) of Electric Components Program is consistent with the program described in NUREG-1801,Section X.E1, Environmental Qualification (EQ) of Electric Components.
Exceptions to NUREG-1801 None
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-53 Enhancements None Operating Experience The Environmental Qualification (EQ) Program at GGNS is routinely audited to ensure that program elements are carried out properly. A program assessment in 2002 revealed no major problems. Operating experience reports were reviewed to determine EQ program impacts for GGNS. This effort indicated no 10 CFR Part 21 issues impacting the EQ program. Areas of weaknesses were addressed and follow-up actions were assigned to improve program effectiveness.
A program assessment in 2009 verified compliance with 10 CFR 50.49 and measured the effectiveness of the program by evaluating the program's overall infrastructure, records, documentation, program implementation, support personnel qualification and knowledge, and the program's impact on plant equipment. Areas of weaknesses were addressed and follow-up actions were assigned to improve program effectiveness.
Providing the proper requirements for the environmental qualification of electrical equipment important to safety, along with the identification of qualified life and specific maintenance/
installation requirements, ensures that the program will remain effective for managing aging effects.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The Environmental Qualification (EQ) of Electric Components Program has been effective at managing aging effects by maintaining equipment within its qualification basis. The Environmental Qualification (EQ) of Electric Components Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-54 B.1.18 EXTERNAL SURFACES MONITORING Program Description The External Surfaces Monitoring Program is an existing program that manages aging effects through visual inspection of external surfaces for evidence of loss of material, cracking and change in material properties. Physical manipulation to detect hardening or loss of strength for elastomers and polymers is also used.
NUREG-1801 Consistency The External Surfaces Monitoring Program, with enhancement, is consistent with the program described in NUREG-1801,Section XI.M36, External Surfaces Monitoring of Mechanical Components.
Exceptions to NUREG-1801 None Enhancements The following enhancements will be implemented prior to the period of extended operation.
Elements Affected Enhancement 3.
Parameters Monitored or Inspected 4.
Detection of Aging Effects 5.
Monitoring The External Surfaces Monitoring Program will be enhanced to include instructions for monitoring aging effects for flexible polymeric components through manual or physical manipulation of the material, with a sample size for manipulation of at least 10 percent of available surface area.
- 4. Detection of Aging Effects The External Surfaces Monitoring Program will be enhanced as follows.
1.
Underground components within the scope of this program will be clearly identified in program documents. Underground components are those for which access is physically restricted.
(cont.)
2.
Instructions will be provided for inspecting all underground components within the scope of this program during each five-year period, beginning ten years prior to the entry into the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-55 Operating Experience During a walkdown of the standby service water "A" pump house in 2009, a loss of coating was discovered on the tailpipe of a discharge safety relief valve, resulting in surface corrosion. The corrosion was determined not to affect the ability of the piping to perform its intended function.
The piping was repaired using the normal work process.
During replacement of heat trace for fire water system piping in 2006, pitting corrosion was detected on the piping. The amount of pitting was determined not to threaten the integrity of the piping. The piping was repaired using the normal work process.
A build-up of scale and rust was found during inspection of a standby service water valve bonnet.
The rust and scale were removed during the inspection, and the valve was found to be fully capable of performing all intended functions.
These examples of the identification of degradation and initiation of corrective action prior to loss of intended function provide evidence that the program is effective for managing aging effects for passive components.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The External Surfaces Monitoring Program has been effective at managing aging effects. The External Surfaces Monitoring Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-56 B.1.19 FATIGUE MONITORING Program Description The Fatigue Monitoring Program is an existing program that ensures that fatigue usage remains within allowable limits by (a) tracking the number of critical thermal and pressure transients for selected components, (b) verifying that the severity of monitored transients are bounded by the design transient definitions for which they are classified, and (c) assessing the impact of the reactor coolant environment on a set of sample critical components.
NUREG-1801 Consistency The Fatigue Monitoring Program, with enhancements, is consistent with the program described in NUREG-1801,Section X.M1, Fatigue Monitoring, with one exception.
Exceptions to NUREG-1801 The Fatigue Monitoring Program is consistent with the program described in NUREG-1801,Section X.M1, Fatigue Monitoring, with the following exception.
Elements Affected Exception 7.
Corrective Actions NUREG-1801 recommends use of a design code limit for cumulative usage factors (CUFs). GGNS applies a more stringent design limit of 0.1 CUFs at high energy line break (HELB) locations. Also, GGNS includes an additional corrective action to evaluate the HELB analysis to address a HELB exclusion location with a CUF that increases to greater than 0.1.1
- 1. The use of a 0.1 limit for CUF at HELB locations is consistent with the criteria stated in UFSAR Section 3.6A.2. Evaluation of the HELB analysis is an additional valid corrective action to address HELB exclusion locations with a CUF that increases to greater than 0.1.
Exception Note:
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-57 Enhancement The following enhancements will be implemented at least two years prior to entering the period of extended operation.
Operating Experience An assessment of the program in 2003 found it to be effective in collecting plant operational data required for the calculation of fatigue usage factors. Data collected and trended through 1999 confirmed that the number of cycles was not trending toward exceeding the allowable number of cycles. This program assessment included actions and recommendations to upgrade the program to enhance its effectiveness.
Analysis of an event in 2003 in which the bottom head cooldown rate limit was exceeded following a plant SCRAM indicated that calculated additional usage factors were small in magnitude. The cumulative usage factors were well within the ASME Section III Code allowable of 1.0 and therefore acceptable.
Elements Affected Enhancement
- 1. Scope of Program A review of the GGNS high energy line break analyses and the corresponding tracking of associated cumulative usage factors will be performed to ensure that the GGNS program adequately manages fatigue usage for these locations.
- 1. Scope of Program Fatigue usage calculations that consider the effects of the reactor water environment will be developed for a set of sample reactor coolant system components. This sample set will include the locations identified in NUREG/CR-6260 and additional plant-specific component locations in the reactor coolant pressure boundary if they are found to be more limiting than those considered in NUREG/CR-6260.
Fen factors will be determined using the formulae sets listed in Section 4.3.3.
- 4. Detection of Aging Effects The GGNS program will be enhanced to revise program documents to provide updates of the fatigue usage calculations on an as-needed basis if an allowable cycle limit is approached, or in a case where a transient definition has been changed, unanticipated new thermal events are discovered, or the geometry of components has been modified.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-58 During a 2011 review of GGNS Class 1 fatigue analyses, deficiencies were identified with fatigue monitoring program activities and documentation. Additional program documentation was obtained, and further corrective actions are being completed under the corrective action program. The correction of the identified program deficiencies provides assurance that the program will be effective for managing the effects of aging due to mechanical fatigue on affected components.
A 2011 study of GGNS plant data showed that the number of plant transients to date were within their design allowable limits.
Operating experience shows that the Fatigue Monitoring Program has been effective in managing aging effects.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The Fatigue Monitoring Program has been demonstrated to maintain the validity of the fatigue design basis for reactor coolant system components designed to withstand the effects of cyclic loads due to reactor system transients.
The Fatigue Monitoring Program assures the fatigue design basis is maintained such that applicable components will continue to perform their intended function consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-59 B.1.20 FIRE PROTECTION Program Description The Fire Protection Program is an existing program that manages cracking, loss of material, and change in material properties through visual inspection of components and structures with a fire barrier intended function. It also manages loss of material for the CO2 and Halon fire suppression systems through periodic visual inspection and testing.
NUREG-1801 Consistency The Fire Protection Program, with enhancements, is consistent with the program described in NUREG-1801,Section XI.M26, Fire Protection.
Exceptions to NUREG-1801 None Enhancements The following enhancements will be implemented prior to the period of extended operation.
Elements Affected Enhancements
- 3. Parameters Monitored or Inspected
- 4. Detection of Aging Effects The Fire Protection Program will be enhanced to require visual inspections of the Halon/CO2 fire suppression system at least once every fuel cycle to examine for signs of corrosion.
- 4. Detection of Aging Effects The Fire Protection Program will be enhanced to require visual inspections of fire damper framing at least once every fuel cycle to check for signs of degradation.
- 4. Detection of Aging Effects The Fire Protection Program will be enhanced to require visual inspections of concrete curbs, manways, hatches, manhole covers, hatch covers, and roof slabs at least once every fuel cycle to confirm that aging effects are not occurring.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-60 Operating Experience In 2003, 2006, 2008, and 2010, QA audits of the Fire Protection Program were conducted.
Results of these audits indicated that the program was effective in meeting intended results. The effectiveness was supported through the results of observations and documentation reviews.
Specific walkdowns and inspections of fire barriers conducted in numerous areas of the plant revealed no significant issues. These audits included actions and recommendations to upgrade the program to enhance its effectiveness.
In 2004, 2005, 2006, and 2007, self-assessments were conducted. The program was assessed from a design perspective, as well as from the operational status and material condition perspective. These assessments also evaluated GGNS documentation and supporting conclusions in the Fire Hazards Analysis/Safe Shutdown Analysis. These self assessments included actions and recommendations to upgrade the program to ensure its effectiveness.
Reviews against established program standards provide assurance that the program will remain effective for managing loss of material of components.
During a walkdown in 2005 to examine fireproofing material, several small pieces of fire proofing material were found to be missing on one side of a vertical steel beam. The fire proofing was repaired.
Audit Inspections were conducted in 2003 through 2010 to ensure the integrity and availability of the Halon systems and low pressure CO2 systems. These inspections included actions and recommendations to upgrade the program to enhance its effectiveness with no significant findings or issues noted. These inspections confirm that the program is effective for managing aging effects for passive components.
In 2005 and 2008, the NRC performed triennial fire protection program inspections. Walkdowns of numerous areas of the plant to assess the material condition of fire protection features were completed with no significant findings or issues noted. These walkdowns confirm that the program is effective for managing aging effects for passive components.
Inspections in 2006, 2007, and 2010 found small tears in the door skin for three fire doors as well as rust at the bottom of a third fire door and cracks in four other fire doors. During the course of fire penetration checks in 2009, a fire barrier penetration was found to be degraded due to missing kaowool. Repairs were performed to correct these conditions.
Identification of program deficiencies, and subsequent corrective actions, provide assurance that the program will remain effective for managing loss of material of components.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-61 Conclusion The Fire Protection Program has been effective at managing aging effects. The Fire Protection Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-62 B.1.21 FIRE WATER SYSTEM Program Description The Fire Water System Program is an existing program that manages loss of material and fouling for components in fire protection systems using preventive, inspection, and monitoring activities, including periodic full-flow flush tests and testing or replacement of sprinkler heads. Applicable industry standards and guidance documents are used to delineate the program.
NUREG-1801 Consistency The Fire Water System Program, with enhancements, is consistent with the program described in NUREG-1801,Section XI.M27, Fire Water System Exceptions to NUREG-1801 None Enhancements The following enhancements will be implemented prior to the period of extended operation.
Elements Affected Enhancements
- 3. Parameters Monitored or Inspected
- 6. Acceptance Criteria The Fire Water System Program will be enhanced to include periodic visual inspection of spray and sprinkler system internals for evidence of degradation. Acceptance criteria will be enhanced to verify no unacceptable degradation.
4.
Detection of Aging Effects 6.
Acceptance Criteria The Fire Water System Program will be enhanced to include periodic inspection of hose reels for degradation. Acceptance criteria will be enhanced to verify no unacceptable degradation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-63 4.
Detection of Aging Effects (cont.)
The Fire Water System Program will be enhanced to include one of the following options.
(1) Wall thickness evaluations of fire protection piping using non-intrusive techniques (e.g.,
volumetric testing) to identify evidence of loss of material will be performed prior to the period of extended operation and periodically thereafter. Results of the initial evaluations will be used to determine the appropriate inspection interval to ensure aging effects are identified prior to loss of intended function.
OR (2) A visual inspection of the internal surface of fire protection piping will be performed upon each entry to the system for routine or corrective maintenance. These inspections will be capable of evaluating (a) wall thickness to ensure against catastrophic failure and (b) the inner diameter of the piping as it applies to the design flow of the fire protection system. Maintenance history shall be used to demonstrate that such inspections have been performed on a representative number of locations prior to the period of extended operation. A representative number is 20% of the population (defined as locations having the same material, environment, and aging effect combination) with a maximum of 25 locations. Additional inspections will performed as needed to obtain this representative sample prior to the period of extended operation.
Elements Affected Enhancements
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-64 Operating Experience Sprinkler systems functional tests were performed in 2004, 2006, and 2008 with no significant discrepancies noted.
Yard hydrant flow checks were performed in 2006, 2008, 2009 and 2010 with no discrepancies noted. The hydrant flow checks done in 2005 and 2007 noted a valve that was stuck closed and hose isolation valve hand wheel was broken. The valve and hand wheel were repaired and returned to service.
During a 2008 inspection, nozzle blockages were discovered in the deluge system for the main transformer. The nozzles were cleaned and inspected for debris and returned to service.
Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing aging effects.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
4.
Detection of Aging Effects 6.
Acceptance Criteria The Fire Water System Program will be enhanced to include a visual inspection of a representative number of locations on the interior surface of below grade fire protection piping at a frequency of at least once every ten years during the period of extended operation.
A representative number is 20% of the population (defined as locations having the same material, environment, and aging effect combination) with a maximum of 25 locations.
Acceptance criteria will be no unacceptable degradation.
4.
Detection of Aging Effects The Fire Water System Program will be enhanced to include testing or replacement of a representative sample of sprinkler heads before the end of the 50-year sprinkler head service life and at 10-year intervals thereafter during the extended period of operation. NFPA-25 defines a representative sample of sprinklers to consist of a minimum of not less than 4 sprinklers or 1 percent of the number of sprinklers per individual sprinkler sample, whichever is greater.
Elements Affected Enhancements
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-65 Conclusion The Fire Water System Program has been effective at managing aging effects. The Fire Water System Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-66 B.1.22 FLOW-ACCELERATED CORROSION Program Description The Flow-Accelerated Corrosion (FAC) Program is an existing program that manages loss of material due to wall thinning for piping and components by conducting appropriate analysis and baseline inspections, determining the extent of thinning, performing follow-up inspections, and taking corrective actions as necessary. The program follows guidelines published by EPRI in NSAC-202L.
NUREG-1801 Consistency The FAC Program, with enhancement, is consistent with the program described in NUREG-1801,Section XI.M17, Flow-Accelerated Corrosion.
Exceptions to NUREG-1801 None Enhancements The following enhancement will be implemented prior to the period of extended operation.
Operating Experience A FAC program analysis in 2001 projected the wall thickness for a condensate system stub tube to be less than the current calculated minimum wall thickness prior to RF13 in 2004. Action was taken to replace the existing carbon steel stub tube prior to reaching this point. Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing aging effects for carbon steel components.
A FAC Program assessment report in 2002 concluded that the program fully met all objectives and no major problem areas or issues were identified. However, several areas for improvement were recognized. This assessment included actions and recommendations to upgrade the program and maintain its effectiveness for managing aging effects.
Elements Affected Enhancement 7.
Corrective Actions The Flow-Accelerated Corrosion Program will be enhanced to revise program documentation to specify that downstream components are monitored closely to mitigate any increased wear when susceptible upstream components are replaced with resistant materials, such as high Cr material.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-67 Evaluation of RF14 FAC wall thickness data in 2005 concluded that all items inspected were acceptable for continued service beyond RF15. A re-inspection index (projected life) was calculated for each of the items.
Evaluation of RF16 FAC wall thickness data in 2008 concluded that all items inspected were acceptable for continued service beyond RF17, with three exceptions. Condition reports were issued for each of these three items and each were repaired. Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing aging effects for carbon steel components.
A FAC Program assessment report in 2009 concluded that the "program health" was good. Data and analyses were found to be handled properly and in accordance with Entergy fleet procedures and industry guidelines. This assessment included actions and recommendations to upgrade the program and maintain its effectiveness for managing aging effects. Identification of program deficiencies, and subsequent corrective actions, provide assurance that the program will remain effective for managing loss of material of components.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The Flow-Accelerated Corrosion Program has been effective at managing aging effects. The Flow-Accelerated Corrosion Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-68 B.1.23 INSERVICE INSPECTION Program Description The Inservice Inspection Program is an existing program that manages aging effects for ASME Class 1, 2, and 3 pressure-retaining components including welds, pump casings, valve bodies, integral attachments and pressure-retaining bolting using volumetric, surface, or visual examination as specified in ASME Section XI code. Every ten years this program is updated to the latest ASME Section XI code edition and addendum approved by the NRC in 10 CFR 50.55a.
NUREG-1801 Consistency The Inservice Inspection Program is consistent with the program described in NUREG-1801,Section XI.M1, ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD.
Exceptions to NUREG-1801 None Enhancements None Operating Experience ISI Program summary reports between 2004 and 2010 reveal compliance (including evaluation or repair of indications/flaws) and provide evidence that the program is effective for managing aging effects in accordance with the ASME Boiler Pressure Vessel Code Section XI.
Self assessments performed in 2005, 2006, 2007, 2010 concluded that ISI program activities are being performed in accordance with ASME Section XI and Grand Gulf program requirements.
From the reviews it was found that GGNS has an appropriate threshold for entering issues in the corrective action program and has taken appropriate corrective actions. Additionally, a similar review of operating experience reports identified that applicable operating experience was being effectively evaluated and implemented.
The ISI Program detects aging effects via visual, surface and ultrasonic inspection. Identification of program deficiencies, and subsequent corrective actions, provide assurance that the program will remain effective for managing loss of material of components. Ultrasonic inspection methods are subject to the performance demonstration requirements of ASME Section XI, Appendix VIII.
In addition, the ISI programs are based on industry-wide experience.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-69 Conclusion The Inservice Inspection Program has been effective at managing aging effects. The Inservice Inspection Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-70 B.1.24 INSERVICE INSPECTION - IWF Program Description The Inservice Inspection - IWF Program is an existing program that manages aging effects for ASME Class 1, 2, 3 and component supports. The scope of inspection for component supports is based on sampling of piping supports and 100% of component supports other than piping as specified in Table IWF-2500-1.
NUREG-1801 Consistency The Inservice Inspection - IWF Program, with enhancements, is consistent with the program described in NUREG-1801,Section XI.S3, ASME Section XI, Subsection IWF.
Exceptions to NUREG-1801 None Enhancements The following enhancements will be implemented prior to the period of extended operation.
Elements Affected Enhancements
- 1. Scope of Program The ISI-IWF Program will be enhanced to address inspections of accessible sliding surfaces.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-71
- 3. Parameters Monitored or Inspected The ISI-IWF Program will be enhanced to clarify that parameters monitored or inspected will include corrosion; deformation; misalignment of supports; missing, detached, or loosened support items; improper clearances of guides and stops; and improper hot or cold settings of spring supports and constant load supports.
Accessible areas of sliding surfaces will be monitored for debris, dirt, or indications of excessive loss of material due to wear that could prevent or restrict sliding as intended in the design basis of the support. Elastomeric vibration isolation elements will be monitored for cracking, loss of material, and hardening.
Structural bolts will be monitored for corrosion and loss of integrity of bolted connections due to self-loosening and material conditions that can affect structural integrity. High-strength structural bolting (actual measured yield strength greater than or equal to 150 ksi or 1,034 MPa in sizes greater than 1 inch nominal diameter) susceptible to stress corrosion cracking (SCC) will be monitored for SCC.
4.
Detection of Aging Effects The ISI-IWF Program will be enhanced to clarify that detection of aging will include the following:
(a) Structural bolting (ASTM A-325, ASTM F1852, and ASTM A490 bolts) and anchor bolts will be monitored for loss of material, loose or missing nuts, loss of pre-load, and cracking of concrete around the anchor bolts.
(b) Volumetric examination comparable to that of ASME Code Section XI, Table IWB-2500-1, Examination Category B-G-1 should be performed for high strength structural bolting to detect cracking in addition to the VT-3 examination. This volumetric examination may be waived with adequate plant-specific justification.
Elements Affected Enhancements
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-72 Operating Experience Results of ISI examinations for pipe hanger and supports for the reactor core isolation cooling system during RF16 in 2008 were acceptable.
Results of ISI examinations for component supports for the reactor recirculation system during RF17 in 2010 were acceptable.
The Inservice Inspection - IWF Program detects aging effects using visual techniques to detect and characterize flaws. These techniques are widely used and have been demonstrated effective at detecting aging effects during inspections performed to meet ASME Section XI Code requirements.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The Inservice Inspection - IWF Program has been effective at managing aging effects. The Inservice Inspection - IWF Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
6.
Acceptance Criteria The ISI-IWF program will be enhanced to include the following as unacceptable conditions.
- Loss of material due to corrosion or wear, which reduces the load bearing capacity of the component support.
- Debris, dirt, or excessive wear that could prevent or restrict sliding of the sliding surfaces as intended in the design basis of the support.
- Cracked or sheared bolts, including high strength bolts, and anchors.
Elements Affected Enhancements
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-73 B.1.25 INSPECTION OF OVERHEAD HEAVY LOAD AND LIGHT LOAD (RELATED TO REFUELING) HANDLING SYSTEMS Program Description The Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems Program is an existing program that consists of periodic inspections and preventive maintenance to manage loss of material of cranes and hoists, based on applicable industry standards and guidance documents. The activities rely on visual examinations and functional testing to ensure that cranes and hoists are capable of sustaining their rated loads, thus ensuring their intended function is maintained during the period of extended operation.
NUREG-1801 Consistency The Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems Program, with enhancements, is consistent with the program described in NUREG-1801,Section XI.M23, Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems.
Exceptions to NUREG-1801 None Enhancements The following enhancements will be implemented prior to the period of extended operation.
Elements Affected Enhancements
- 1. Scope of Program The program will be enhanced to include monitoring rails in the rail system for the aging effect of wear and structural connections/bolting for loose or missing bolts, nuts, pins or rivets.
Additionally, the program will be clarified to include visual inspection of structural components and structural bolts for loss of material due to various mechanisms and structural bolting for loss of preload due to self-loosening.
- 6. Acceptance Criteria Acceptance criteria will be revised to state that any significant loss of material for structural components and structural bolts, and significant wear of rails in the rail system, is evaluated according to ASME B30.2 or other applicable industry standard in the ASME B30 series.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-74 Operating Experience During a 2008 turbine building crane inspection, the bridge girder junction bolts, the bridge cross-tie bolts and the bridge drive coupling bolts were found to be loose. Corrective actions included tightening of all bolting.
During a 2009 inspection of the polar crane rail clips, a broken stud was found at azimuth 90.
The rail clips were repaired.
Identification of deficiencies, and subsequent corrective actions, provide assurance that the program will remain effective for managing aging effects of components.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems Program has been effective at managing aging effects. The Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-75 B.1.26 INTERNAL SURFACES IN MISCELLANEOUS PIPING AND DUCTING COMPONENTS Program Description The Internal Surfaces in Miscellaneous Piping and Ducting Components Program is a new program that manages the effects of aging using visual inspections of the internal surfaces of piping and components during periodic surveillances or maintenance activities when the surfaces are accessible for visual inspection. Physical manipulation or pressurization to detect hardening or loss of strength for elastomers and polymers is also used.
This program will be implemented prior to the period of extended operation.
NUREG-1801 Consistency The Internal Surfaces in Miscellaneous Piping and Ducting Components Program is consistent with the program described in NUREG-1801,Section XI.M38, Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components.
Exceptions to NUREG-1801 None Enhancements None Operating Experience The Internal Surfaces in Miscellaneous Piping and Ducting Components Program is a new program. Industry operating experience was considered in the development of this program.
Plant operating experience will be gained as the program is implemented and will be factored into the program via the confirmation and corrective action elements of the GGNS 10 CFR 50 Appendix B quality assurance program.
The methods used in this program to detect aging effects are proven industry techniques that have been effectively used at GGNS in other programs. As such, operating experience assures that implementation of the Internal Surfaces in Miscellaneous Piping and Ducting Components Program will manage the effects of aging such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-76 Conclusion The Internal Surfaces in Miscellaneous Piping and Ducting Components Program will be effective for managing aging effects since it will incorporate proven monitoring techniques, acceptance criteria, corrective actions, and administrative controls. The Internal Surfaces in Miscellaneous Piping and Ducting Components Program provides assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-77 B.1.27 MASONRY WALL Program Description The Masonry Wall Program is an existing program that manages aging effects for each masonry wall within the scope of license renewal. The program includes visual inspection of masonry walls including 10 CFR 50.48-required masonry walls, radiation-shielding masonry walls, and masonry walls with the potential to affect safety-related components. Structural steel components of masonry walls are managed by the Structures Monitoring Program. Masonry walls are visually examined at a frequency selected to ensure there is no loss of intended function between inspections.
NUREG-1801 Consistency The Masonry Wall Program, with enhancements, is consistent with the program described in NUREG-1801,Section XI.S5, Masonry Wall Program.
Exceptions to NUREG-1801 None Enhancements The following enhancements will be implemented prior to the period of extended operation.
Elements Affected Enhancements
- 3. Parameters Monitored or Inspected The Masonry Wall Program will be enhanced to clarify that parameters monitored or inspected will include monitoring gaps between the supports and masonry walls that could potentially affect wall qualification.
- 4. Detection of Aging Effects The Masonry Wall Program will be enhanced to clarify that detection of aging effects require masonry walls to be inspected every five years unless technical justification is provided to extend the inspection to a period not to exceed ten years.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-78 Operating Experience Program assessments covering the period from 2001 to 2007 identified no problems with masonry walls. The assessments concluded the structure inspection program is adequate and effective. Reviews against established program standards provide assurance that the program will remain effective for managing loss of material of components.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The Masonry Wall Program has been effective at managing aging effects. The Masonry Wall Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-79 B.1.28 NON-EQ CABLE CONNECTIONS Program Description The Non-EQ Cable Connections Program is a new one-time inspection program that provides reasonable assurance that the intended functions of the metallic parts of electrical cable connections are maintained consistent with the current licensing basis through the period of extended operation. Cable connections included are those connections susceptible to age-related degradation resulting in increased resistance of connection due to thermal cycling, ohmic heating, electrical transients, vibration, chemical contamination, corrosion, or oxidation that are not subject to the environmental qualification requirements of 10 CFR 50.49.
This program provides for one-time inspections that will be completed prior to the period of extended operation on a sample of connections. The factors considered for sample selection will be application (medium and low voltage, defined as < 35 kV), circuit loading (high loading),
connection type, and location (high temperature, high humidity, vibration, etc.). The representative sample size will be based on twenty percent of the connection population with a maximum sample of 25.
The inspections will be performed prior to the period of extended operation.
NUREG-1801 Consistency The Non-EQ Cable Connections Program is consistent with the program described in NUREG-1801,Section XI.E6, Electrical Cable Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Exceptions to NUREG-1801 None Enhancements None Operating Experience The Non-EQ Cable Connections Program is a new program. Industry operating experience was considered in the development of this program. Plant operating experience will be gained as the program is implemented and will be factored into the program via the confirmation and corrective action elements of the GGNS 10 CFR 50 Appendix B quality assurance program.
This inspection program applies to potential aging effects for which there is currently no operating experience at GGNS indicating the need for an aging management program.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-80 The elements of the program inspections (e.g., the scope of the inspections and inspection techniques) are consistent with industry practice and have been used effectively at GGNS in other programs.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The Non-EQ Cable Connections Program will be effective for managing aging effects since it will incorporate proven monitoring techniques, acceptance criteria, corrective actions, and administrative controls. The Non-EQ Cable Connections Program provides assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-81 B.1.29 NON-EQ INACCESSIBLE POWER CABLES (400 V TO 35 KV)
Program Description The Non-EQ Inaccessible Power Cables (400 V to 35 kV) Program is an existing condition monitoring program that manages the aging effects on the following inaccessible power (400 V to 35kV) cable systems.
Cable ID Voltage Level Description Associated Manholes 2DR104B1, B2 34.5 kV ESF XFMR 21 MH22, MH24 1DR104B1, B2 34.5 kV ESF XFMR 11 MH14 2DR102D1, D2, D3 34.5 kV BOP XFMR 12B MH14, MH10 1DR102D1, D2 34.5 kV BOP XFMR 12A MH14, MH10 1DR101D1, D2, D3 34.5 kV BOP XFMR 11B MH23, MH25, MH5 1DR101C1, C2 34.5 kV BOP XFMR 11A MH13, MH25, MH5 1AA5031 4.16 kV SSW Pump A Motor MH20, MH1 1BA6161 4.16 kV SSW Pump B Motor MH21, MH1 1AA5041 4.16 kV Load Center 15BA5 MH20, MH1 1BA6151 4.16 kV Load Center 16BB5 MH21, MH1 1CB701241 1CB701242 480 V HPCS SW Pump Motor P41C002 MH2, MH3 1CB711061 480 V SSW Loop C RTN TO CLG TWR A MOV P41F011 MH2, MH3 1CB701132 480 V HPCS DSL Gen Jacket Water Heater P81B003B MH3 1CB701141 480 V HPCS DSL Gen Space Heater P81S001 MH3 1CB701151 480 V HPCS DSL Soak Back LO CIRC Pump Motor P81C004B MH3 1CB701182 480 V HPCS DSL Gen Jacket Water Heater P81B003A MH3 1CB701191 480 V HPCS DSL Soak Back LO CIRC Pump Motor P81C004A MH3 1CB701212 1CB701213 480 V HPCS DSL Gen Room Outside Air Fan K77C002 MH3
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-82 The Non-EQ Inaccessible Power Cables (400 V to 35 kV) Program includes periodic actions to prevent inaccessible cables from being exposed to significant moisture. In this program, inaccessible power (400 V to 35kV) cables exposed to significant moisture will be tested at least once every six years to provide an indication of the condition of the cable insulation properties.
Test frequencies are adjusted based on test results and operating experience. The specific type of test performed is a proven test for detecting deterioration of the cable insulation. The program includes periodic inspections for water accumulation in manholes at least once every year (annually). In addition to the periodic manhole inspections, manhole inspections for water after events such as heavy rain or flooding will be performed. Inspection frequency will be increased as necessary based on evaluation of inspection results.
NUREG-1801 Consistency The Non-EQ Inaccessible Power Cables (400 V to 35kV) Program, with enhancements, is consistent with the program described in NUREG-1801,Section XI.E3, Inaccessible Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements.
Exceptions to NUREG-1801 None Enhancements The following enhancements will be implemented prior to the period of extended operation, and the first cable tests and manhole inspections will be completed prior to the period of extended operation.
1CB711031 480 V HPCS DSL Gen FO Transfer Pump Motor P81C001 MH3 Elements Affected Enhancements
- 1. Scope The Non-EQ Inaccessible Power Cables (400 V to 35kV) Program will be enhanced to include low-voltage (400 V to 2 kV) power cables.
Cable ID Voltage Level Description Associated Manholes
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-83 Operating Experience In the response to GL 2007-01, GGNS reported one cable failure, which was a 480V cable to a motor driven fire pump. The apparent cause was listed as age degradation, moisture, and possible damage during construction that accelerated or created susceptibility to moisture.
GGNS has no record of other failures for these types of cables in an underground application.
Based on this information and other industry operating experience, Entergy created a fleet cable reliability program that was effective 12/31/2009 for GGNS. The purpose of the program is to provide the means to effectively manage underground medium voltage cables to achieve high reliability while reducing the likelihood of in-service failures. Prior to and after the program effective date, operating experience was used to assist with the program implementation.
During a manhole internal inspection, three electrical manholes were found to be full of water.
Two of the manholes had inoperable sump pumps as identified by previous level switch functional PM tasks. The other manhole had no installed sump pump. For the remaining manholes, there was no indication of problems with the sump pumps in these manholes.
Corrective actions included identifying which electrical manholes contain sump pumps and which electrical manholes do not. Corrective actions were established to periodically inspect and pump down the manholes with no sump pumps installed.
The standby service water motors are periodically meggered, including the 4160 V cables that are routed through manholes without sump pumps installed. This provides cable monitoring that ensures that no significant degradation exists at the time the megger is performed.
- 2. Preventive Actions The Non-EQ Inaccessible Power Cables (400 V to 35kV) Program will be enhanced. Condition-based inspections of manholes not automatically dewatered by a sump pump will be performed following periods of heavy rain or potentially high water table conditions, as indicated by river level.
The Non-EQ Inaccessible Power Cables (400 V to 35kV) Program will be clarified that the manhole inspections will include direct observation that cables are not wetted or submerged, that cables/splices and cable support structures are intact, and verification that dewatering/drainage systems (i.e., sump pumps) and associated alarms if applicable operate properly.
Elements Affected Enhancements
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-84 The fleet cable reliability program provides guidance relative to ensuring adequate inspection criteria for cables contained in manholes and includes instructions for inspection of these cables and any required action to initiate further actions to mitigate water intrusion and/or increase the frequency of inspection and dewatering activities. Continued inspection of manholes for water intrusion is providing further information for inspection and testing frequency.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The Non-EQ Inaccessible Power Cables (400 V to 35kV) Program has been effective at managing aging effects. The Non-EQ Inaccessible Power Cables (400 V to 35kV) Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-85 B.1.30 NON-EQ INSTRUMENTATION CIRCUITS TEST REVIEW Program Description The Non-EQ Instrumentation Circuits Test Review Program is a new performance monitoring program that will manage the aging effects of the applicable cables in the following systems or sub-systems.
Neutron monitoring - intermediate range channels (IRMs)
Neutron monitoring - local power range monitors (LPRMs)
Neutron monitoring - average power range monitors (APRMs)1 Process radiation monitoring:
Main steam radiation monitoring
Containment and drywell ventilation exhaust monitoring
Fuel handling area exhaust and the fuel handling area pool sweep exhaust monitoring
Control room ventilation monitoring The Non-EQ Instrumentation Circuits Test Review Program assures the intended functions of sensitive, high-voltage, low-signal cables exposed to adverse localized equipment environments caused by heat, radiation and moisture (i.e., neutron flux monitoring instrumentation and process radiation monitoring) can be maintained consistent with the current licensing basis through the period of extended operation. Most sensitive instrumentation circuit cables and connections are included in the instrumentation loop calibration at the normal calibration frequency, which provides sufficient indication of the need for corrective actions based on acceptance criteria related to instrumentation loop performance. The review of calibration results will be performed once every ten years, with the first review occurring before the period of extended operation.
For sensitive instrumentation circuit cables that are disconnected during instrument calibrations, testing using a proven method for detecting deterioration for the insulation (such as insulation resistance tests or time domain reflectometry) will occur at least once every ten years, with the first test occurring before the period of extended operation. Applicable industry standards and guidance documents are used to delineate the program.
This program will be implemented prior to the period of extended operation.
NUREG-1801 Consistency The program will be consistent with the program described in NUREG-1801,Section XI.E2, Insulation Material for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-86 Exceptions to NUREG-1801 None Enhancements None Operating Experience The Non-EQ Instrumentation Circuits Test Review Program is a new program. Industry operating experience will be considered when implementing this program. Plant operating experience for this program will be gained as it is implemented during the period of extended operations and will be factored into the program via the confirmation and corrective action elements of the GGNS 10 CFR 50 Appendix B quality assurance program.
This inspection program applies to potential aging effects for which there is currently no operating experience at GGNS indicating the need for an aging management program. A search of GGNS operating experience identified no age-related failures of neutron monitoring and high range radiation monitoring system cables and connections at GGNS, and no aging mechanisms not considered in NUREG-1801 have been identified. The elements of the program inspections (e.g., the scope of the inspections and inspection techniques) are consistent with industry practice and have been used effectively at GGNS in other programs.
The GGNS program is based on the program description in NUREG-1801, which in turn is based on industry operating experience that demonstrates that this program is effective for managing the aging effects described herein. As such, operating experience assures that implementation of the Non-EQ Instrumentation Circuits Test Review Program will manage the effects of aging such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The Non-EQ Instrumentation Circuits Test Review Program will be effective for managing aging effects since it will incorporate proven monitoring techniques, acceptance criteria, corrective actions, and administrative controls. The Non-EQ Instrumentation Circuits Test Review Program provides assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-87 B.1.31 NON-EQ INSULATED CABLES AND CONNECTIONS Program Description The Non-EQ Insulated Cables and Connections Program is a new condition monitoring program that assures the intended functions of insulated cables and connections exposed to adverse localized environments caused by heat, radiation and moisture can be maintained consistent with the current licensing basis through the period of extended operation. An adverse localized environment is a condition in a limited plant area that is significantly more severe than the plant design environment for the cable or connection insulation materials.
A representative sample consisting of accessible insulated cables and connections within the scope of license renewal installed in an adverse localized environment will be visually inspected for cable and connection jacket surface anomalies such as embrittlement, discoloration, cracking or surface contamination. The program sample consists of all accessible cables and connections in localized adverse environments. This program sample of accessible cables will represent, with reasonable assurance, all cables and connections in the adverse localized environment.
This program will visually inspect accessible cables in an adverse localized environment at least once every 10 years, with the first inspection prior to the period of extended operation.
This program will be implemented prior to the period of extended operation.
NUREG-1801 Consistency The Non-EQ Insulated Cables and Connections Program will be consistent with the program described in NUREG-1801,Section XI.E1, Insulation Material for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements.
Exceptions to NUREG-1801 None Enhancements None Operating Experience The Non-EQ Insulated Cables and Connections Program is a new program. Industry operating experience will be considered when implementing this program. Plant operating experience for this program will be gained as it is implemented during the period of extended operations and will be factored into the program via the confirmation and corrective action elements of the GGNS 10 CFR 50 Appendix B quality assurance program.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-88 Cables for a nonsafety-related valve located in the GGNS feedwater heater room were found to have heat-related degradation. This valve and associated cables have no license renewal intended function. A search of GGNS operating experience identified no age-related failures of cables or cable connections with a license renewal intended function at GGNS.
The GGNS program is based on the program description in NUREG-1801, which in turn is based on industry operating experience that demonstrates that this program is effective for managing the aging effects described herein. As such, operating experience assures that implementation of the Non-EQ Insulated Cables and Connections Program will manage the effects of aging such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The Non-EQ Insulated Cables and Connections Program will be effective for managing aging effects since it will incorporate proven monitoring techniques, acceptance criteria, corrective actions, and administrative controls. The Non-EQ Insulated Cables and Connections Program provides assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-89 B.1.32 OIL ANALYSIS Program Description The Oil Analysis Program is an existing program that ensures that loss of material, cracking, and fouling are not occurring by maintaining oil environments free of contaminants (primarily water and particulates). Testing activities include sampling and analysis of lubricating oil.
The One-Time Inspection Program utilizes inspections or non-destructive evaluations of representative samples to verify that the Oil Analysis Program has been effective at managing aging effects.
NUREG-1801 Consistency The Oil Analysis Program, with enhancements, is consistent with the program described in NUREG-1801,Section XI.M39, Lubricating Oil Analysis.
Exceptions to NUREG-1801 None Enhancements The following enhancements will be implemented prior to the period of extended operation.
Operating Experience Division I & II diesel engine lubricating oil sample results for 2001 through 2010 were satisfactory.
Fire water diesel engine lubricating oil sample results for 2000 through 2010 were satisfactory.
However, both fire water diesels had samples with high particle count. This was attributed to faulty lab equipment, and samples were taken at more frequent intervals to track the conditions.
Even though the samples showed high particle count, the level was still below the limit and was evaluated to be satisfactory for continued service.
Elements Affected Enhancements
- 1. Scope of Program The Oil Analysis Program will be enhanced to include piping and components within the main generator system (N41) with an internal environment of lube oil.
- 3. Parameters Monitored or Inspected
- 4. Detection of Aging Effects
- 6. Acceptance Criteria The Oil Analysis Program will be enhanced to provide a formalized analysis technique for particulate counting.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-90 Samples of crankcase oil from the high pressure core spray diesel generator (Division III) taken between 2005 and 2010 revealed possible high metal count. Continued observation was called for to determine if this was caused by metal wear particles or malfunction of lab equipment. From 2008 on the samples were normal, indicating a malfunction of lab equipment in the earlier samples.
A sample of Division I diesel engine lubricating oil in 2007 indicated trace amounts of moisture contamination. Oil viscosity and other parameters were within manufacturer specifications. A contaminated sample tubing was suspected, and a resample was performed with clean sampling apparatus to confirm this.
A sample of drywell purge compressor lubricating oil in 2008 indicated moisture contamination.
A resample was performed and no abnormal results were noted.
Identification of signs of possible degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing aging effects for carbon steel components.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The Oil Analysis Program has been effective at managing aging effects. The Oil Analysis Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-91 B.1.33 ONE-TIME INSPECTION Program Description The One-Time Inspection Program is a new program that consists of a one-time inspection of selected components to accomplish one of the following:
Verify the effectiveness of an AMP that is designed to prevent or minimize aging to the extent that it will not cause the loss of intended function during the period of extended operation.
Confirm the insignificance of an aging effect for situations in which additional confirmation is appropriate.
Inspections that verify unacceptable degradation is not occurring will be used.
The sample size of components to be inspected will be based on an assessment of materials, environment, aging effects, and operating experience. Identification of inspection locations will be based on the potential for the aging effect to occur. Examination techniques will be established NDE methods with a demonstrated history of effectiveness in detecting the aging effect of concern, including visual, ultrasonic, and surface techniques. Acceptance criteria will be based on applicable ASME or other appropriate standards, design basis information, or vendor-specified requirements and recommendations. The need for follow-up examinations will be evaluated.
The program will include activities to verify effectiveness of aging management programs and activities to confirm the insignificance of aging effects as described below.
Diesel fuel monitoring program One-time inspection activity will verify the effectiveness of the diesel fuel monitoring aging management programs by confirming that unacceptable loss of material is not occurring.
Oil analysis program One-time inspection activity will verify the effectiveness of the oil analysis aging management programs by confirming that unacceptable cracking, loss of material, and fouling is not occurring.
Water chemistry control program One-time inspection activity will verify the effectiveness of the water chemistry control
- BWR aging management program by confirming that unacceptable cracking, loss of material, and fouling is not occurring.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-92 The inspection will be performed within the ten years prior to the period of extended operation.
NUREG-1801 Consistency The One-Time Inspection Program will be consistent with the program described in NUREG-1801,Section XI.M32, One-Time Inspection.
Exceptions to NUREG-1801 None Enhancements None Operating Experience The One-Time Inspection Program is a new program. Industry operating experience will be considered when implementing this program. Plant operating experience for this program will be gained as it is implemented and will be factored into the program via the confirmation and corrective action elements of the GGNS 10 CFR 50 Appendix B quality assurance program.
As stated in NUREG-1801,Section XI.M32, the elements of these inspections (e.g., the scope of the inspections and inspection techniques) are consistent with industry practice and use developed and approved industry techniques for inspection such as UT and visual exams. These techniques have also been proven effective for detection of aging effects outside of this program as documented in operating experience for other programs such as Flow-Accelerated Corrosion and ASME Section XI. Accordingly, there is reasonable assurance that this new aging management program will be effective during the period of extended operation.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The One-Time Inspection Program will be effective for managing aging effects since it will incorporate proven monitoring techniques, acceptance criteria, corrective actions, and administrative controls. The One-Time Inspection Program provides assurance that the Water Chemistry Control, Diesel Fuel Monitoring, and Oil Analysis programs will be effective in managing the effects of aging to ensure component intended functions can be maintained in accordance with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-93 B.1.34 ONE-TIME INSPECTION - SMALL-BORE PIPING Program Description The One-Time Inspection - Small-Bore Piping Program is a new program that augments ASME Code,Section XI requirements and is applicable to small-bore ASME Code Class 1 piping and components with a nominal pipe size diameter less than 4 inches (NPS < 4) and greater than or equal to NPS 1, in systems that have not experienced cracking of ASME Code Class 1 small-bore piping. GGNS has not experienced cracking of ASME Code Class 1 small-bore piping due to stress corrosion, cyclical (including thermal, mechanical, and vibration fatigue) loading, or thermal stratification and thermal turbulence. The program can also be used for systems that have experienced cracking but have implemented design changes to effectively mitigate cracking.
This program provides a one-time volumetric inspection of a sample of these Class 1 piping locations that are susceptible to cracking. The program includes pipes, fittings, branch connections, and all full and partial penetration (socket) welds.
This program includes a statistically significant sampling approach. Sample selection is based on susceptibility to stress corrosion, cyclic loading (including thermal, mechanical, and vibration fatigue), or thermal stratification and thermal turbulence.
The program includes measures to verify that degradation is not occurring, thereby either confirming that there is no need to manage aging-related degradation or validating the effectiveness of any existing program for the period of extended operation. If evidence of cracking is revealed by this one-time inspection, follow-up periodic inspection will be managed by a plant-specific program.
The inspection will be performed within the six-year period prior to the period of extended operation.
NUREG-1801 Consistency The One-Time Inspection - Small-Bore Piping Program will be consistent with the program described in NUREG-1801,Section XI.M35, One-Time Inspection of ASME Code Class 1 Small-Bore Piping Program.
Exceptions to NUREG-1801 None Enhancements None
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-94 Operating Experience The One-Time Inspection - Small-Bore Piping Program is a new program. Industry operating experience will be considered when implementing this program. Plant operating experience for this program will be gained as it is implemented during the period of extended operations and will be factored into the program via the confirmation and corrective action elements of the GGNS 10 CFR 50 Appendix B quality assurance program.
This inspection program applies to a potential aging effect (cracking of ASME Code Class 1 piping less than 4 inches nominal pipe size) for which there is no operating experience at GGNS that indicates the need for an aging management program. As stated in NUREG 1801,Section XI.M35, this program will use volumetric or destructive inspection techniques with demonstrated capability and a proven industry record to detect cracking in piping weld and base material. As such, operating experience assures that implementation of the One-Time Inspection - Small-Bore Piping Program will manage the effects of aging such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The One-Time Inspection - Small-Bore Piping Program will be effective for managing aging effects since it will incorporate proven inspection techniques, acceptance criteria, corrective actions, and administrative controls. The One-Time Inspection - Small-Bore Piping Program provides assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-95 B.1.35 PERIODIC SURVEILLANCE AND PREVENTIVE MAINTENANCE Program Description There is no corresponding NUREG-1801 program.
The Periodic Surveillance and Preventive Maintenance Program is an existing program that manages aging effects not managed by other aging management programs, including loss of material, cracking, and change in material properties.
Credit for program activities has been taken in the aging management review of the following systems and structures.
Containment Building Visually inspect and manually flex the rubber gasket/seal for upper containment pool gates to verify the absence of cracks and significant change in material properties.
Low pressure core spray system (LPCS)
Use visual or other NDE techniques to inspect external surface of LPCS piping passing through the waterline region of suppression pool to manage loss of material.
Residual heat removal (RHR) system Use visual or other NDE techniques to inspect external surface of RHR piping passing through the waterline region of suppression pool to manage loss of material.
Pressure relief system Use visual or other NDE techniques to inspect external surface of pressure relief system piping passing through the waterline region of the suppression pool to manage loss of material.
Reactor core isolation cooling (RCIC) system Use visual or other NDE techniques to inspect external surfaces of RCIC system piping passing through the waterline region of the suppression pool to manage loss of material.
Nonsafety-related systems affecting safety-related systems Visually inspect the internal surfaces of a representative sample of piping in the control rod drive (CRD) system to manage loss of material.
Visually inspect the internal surfaces of a representative sample of piping and valve bodies in the circulating water system (N71) to manage loss of material.
Visually inspect the internal surfaces of a representative sample of piping and valve bodies in the floor and equipment drain system (P45) to manage loss of material.
High pressure core spray (HPCS) system Use visual or other NDE techniques to inspect HPCS piping passing through the waterline region of the suppression pool to manage loss of material.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-96 Evaluation 1.
Scope of Program The Periodic Surveillance and Preventive Maintenance Program, with regard to license renewal, includes the specific structures and components identified in the aging management reviews as listed in the table above.
2.
Preventive Actions Similar to other condition monitoring programs described in NUREG-1801, the Periodic Surveillance and Preventive Maintenance Program does not include preventive actions.
3.
Parameters Monitored/Inspected The GGNS Periodic Surveillance and Preventive Maintenance Program monitors and inspects parameters linked to the degradation of the particular structure or component-intended function(s).
4.
Detection of Aging Effects Preventive maintenance activities and periodic surveillances provide for periodic component inspections to detect aging effects. Inspection intervals are established such that they provide timely detection of degradation prior to loss of intended functions.
Inspection intervals, sample sizes, and data collection methods are dependent on component material and environment and take into consideration industry and plant-specific operating experience and manufacturers' recommendations.
Established techniques such as visual inspections are used. Each inspection occurs at least once every five years.
For each activity that refers to a representative sample, a representative sample is 20%
of the population (defined as components having the same material, environment, and aging effect combination) with a maximum of 25 components.
Floor and equipment drain system Use visual or other NDE techniques to inspect piping below the waterline in the in-scope sumps to manage loss of material.
Visually inspect the internal surfaces of a representative sample of piping, drain housings, and valve bodies in the floor and equipment drain system (P45) to manage loss of material.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-97 5.
Monitoring and Trending Preventive maintenance activities provide for monitoring and trending of aging degradation. Inspection intervals are established such that they provide for timely detection of component degradation. Inspection intervals are dependent on component material and environment and take into consideration industry and plant-specific operating experience and manufacturers' recommendations.
6.
Acceptance Criteria Periodic Surveillance and Preventive Maintenance Program acceptance criteria are defined in specific inspection procedures. The procedures confirm that the structure or component intended function(s) are maintained by verifying the absence of aging effects or by comparing applicable parameters to limits established by plant design basis.
7.
Corrective Actions Corrective actions, including root cause determination and prevention of recurrence, are implemented in accordance with requirements of 10 CFR Part 50, Appendix B.
8.
Confirmation Process This element is discussed in Section B.0.3.
9.
Administrative Controls This element is discussed in Section B.0.3.
- 10. Operating Experience Typical inspection results of this program include the following.
NDE measurements were made on Division II diesel generator exhaust piping in 2005 to check wall thickness. Analysis of the data showed acceptable results. There was no other indication of aging such as erosion or corrosion. Preventive maintenance test results confirming the absence of significant wall loss provides evidence that the program is effective for managing loss of material.
In 2006, visual inspection of the internal surfaces of a check valve in the component cooling water system found significant wear. The affected parts were replaced and the valve was returned to service. There was no other indication of aging such as erosion or corrosion. Identification of signs of possible degradation and corrective action prior to loss of intended function provides evidence that the program is effective for managing aging effects for components.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-98 The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Enhancements The following enhancements will be implemented prior to the period of extended operation.
Conclusion The Periodic Surveillance and Preventive Maintenance Program has been effective at managing aging effects. The Periodic Surveillance and Preventive Maintenance Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Elements Affected Enhancements
- 1. Scope of Program
- 3. Parameters Monitored or Inspected
- 4. Detection of Aging Effects
- 6. Acceptance Criteria The Periodic Surveillance and Preventive Maintenance Program will be enhanced to revise program guidance documents as necessary to include all activities described in the table provided in the program description.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-99 B.1.36 PROTECTIVE COATING MONITORING AND MAINTENANCE Program Description The Protective Coating Monitoring and Maintenance Program is an existing program that monitors and maintains service level I coatings inside containment. The program assesses coating condition through visual inspections.
NUREG-1801 Consistency The Protective Coating Monitoring and Maintenance Program, with enhancements, is consistent with the program described in NUREG-1801,Section XI.S8, Protective Coating Monitoring and Maintenance Program.
Exceptions to NUREG-1801 None Enhancements The following enhancements will be implemented prior to the period of extended operation.
Operating Experience Visual inspection of coatings in 2000, 2002, 2005, 2007, 2008, and 2010 found no conditions that required immediate repair. Locations with varying degrees of corrosion and minor pitting were noted for future inspections.
Elements Affected Enhancements
- 3. Parameters Monitored or Inspected The Protective Coating Monitoring and Maintenance Program will be enhanced to include parameters monitored or inspected per the guidance provided in ASTM D5163-08.
- 4. Detection of Aging Effects The Protective Coating Monitoring and Maintenance Program will be enhanced to provide for inspection of coatings near sumps or screens associated with the emergency core cooling system.
- 6. Acceptance Criteria The Protective Coating Monitoring and Maintenance Program will be enhanced to include acceptance criteria per ASTM D 5163-08.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-100 The Protective Coating Monitoring and Maintenance Program detects aging effects using visual techniques to detect and characterize flaws. These techniques are widely used and have been demonstrated effective at detecting aging effects during inspections performed for other programs at GGNS. The application of these proven methods provides assurance that the effects of aging will be managed such that the Protective Coating Monitoring and Maintenance Program components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The Protective Coating Monitoring and Maintenance Program has been effective at managing aging effects. The Protective Coating Monitoring and Maintenance Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-101 B.1.37 REACTOR HEAD CLOSURE STUDS Program Description The Reactor Head Closure Studs Program is an existing program that manages cracking and loss of material for reactor head closure stud bolting using inservice inspection and preventive measures. ASME Section XI examination and inspection requirements specified in Table IWB-2500-1 are used. The program also relies on recommendations to address reactor head closure stud bolting degradation listed in NUREG-1339 and NRC Regulatory Guide 1.65.
NUREG-1801 Consistency The Reactor Head Closure Studs Program is consistent with the program described in NUREG-1801,Section XI.M3, Reactor Head Closure Stud Bolting, with one exception.
Exceptions to NUREG-1801 The Reactor Head Closure Studs Program is consistent with the program described in NUREG-1801,Section XI.M3, Reactor Head Closure Stud Bolting, with the following exception.
Elements Affected Exception
- 2. Preventive Actions NUREG-1801 recommends use of bolting material for closure studs that has an actual measured yield strength less than 1,034 megapascals (MPa) (150 kilo-pounds per square inch). GGNS uses bolting material for closure studs with a maximum reported ultimate tensile strength below 170 kilo-pounds per square inch.1
- 1. The criterion of actual yield strength less than 150 kilo-pounds per square inch (ksi) was recommended in Section 3 of NUREG-1339 to be used as the level for consideration of vulnerability to stress corrosion cracking (SCC). The studs, nuts and washers at GGNS are fabricated from SA 540 Grade B23 or B24 carbon steel, which has a minimum yield strength of 130 ksi. Data relative to actual yield strength for the installed reactor head closure studs is not available. However, SA 540 Grades B23 and B24 are high-strength, low alloy materials that, when tempered to a maximum tensile strength of 170 ksi, are relatively immune to stress corrosion cracking. Therefore, the studs installed at GGNS are relatively immune to stress corrosion cracking. Nevertheless, since the actual yield strength of the installed studs is not known, the aging management review conservatively identified the stud material as "high strength low alloy steel" susceptible to the aging effect of cracking. The examination methods used for stud inspection in the Reactor Head Closure Studs Program are appropriate to identify cracking. Therefore, the 150 ksi actual yield strength preventive measure is not necessary to assure that the reactor head closure studs can perform their intended function consistent with the current licensing basis through the period of extended operation.
Exception Note
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-102 Enhancements None Operating Experience Surface examination of reactor pressure vessel (RPV) studs, nuts, and washers during RF11 and RF12 from 2001 through 2010 identified no relevant indications. Continuing examination of the studs, washers, and nuts and evaluation of the results provide evidence that the program remains effective in managing and detecting cracking and loss of material in the bolting.
The Reactor Head Closure Studs Program detects aging effects using NDE visual, surface and volumetric techniques to detect and characterize flaws. These techniques are widely used and have been demonstrated effective at detecting aging effects during inspections performed to meet ASME Section XI Code requirements. The application of these proven methods provides assurance that the effects of aging will be managed such that the Reactor Head Closure Studs Program components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The Reactor Head Closure Studs Program has been effective at managing aging effects. The Reactor Head Closure Studs Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-103 B.1.38 REACTOR VESSEL SURVEILLANCE Program Description The Reactor Vessel Surveillance Program is an existing program that manages reduction of fracture toughness for reactor vessel beltline materials using material data and dosimetry. The program includes all reactor vessel beltline materials as defined by 10 CFR 50 Appendix G, Section II.F, and complies with 10CFR50, Appendix H for vessel material surveillance. An integrated surveillance program based on staff-approved BWRVIP documents (including BWRVIP-86-A, BWRVIP-102, BWRVIP-135) has been approved for use by NRC.
NUREG-1801 Consistency The Reactor Vessel Surveillance Program, with enhancements, is consistent with the program described in NUREG-1801,Section XI.M31, Reactor Vessel Surveillance.
Exceptions to NUREG-1801 None Enhancements The following enhancement will be implemented prior to the period of extended operation.
Operating Experience GGNS has committed to using the Boiling Water Reactor Vessel and Internals Project (BWRVIP)
Integrated Surveillance Program (ISP). The fact that the plant participates in the BWRVIP ISP ensures that future operating experience from all participating BWRs will be factored into this program.
Elements Affected Enhancements
- 5. Monitoring and Trending The GGNS Reactor Vessel Surveillance Program will be enhanced to ensure that the additional requirements specified in the final NRC safety evaluation for BWRVIP-86 Revision 1 will be addressed before the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-104 Updated values for vessel fluence and applicable surveillance capsule materials data were used for the 2005 publication of the calculation on which the management of reactor vessel fracture toughness is based. Updated vessel surveillance capsule material test results and data from the BWRVIP were used for a revision of this calculation in 2008, to provide the updated adjusted reference temperature values and to evaluate impact on the pressure-temperature curves. The analysis in this calculation confirmed that the aging effect of reduction of fracture toughness is adequately managed.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4 Conclusion The Reactor Vessel Surveillance Program has been effective at managing aging effects. The Reactor Vessel Surveillance Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-105 B.1.39 RG 1.127, INSPECTION OF WATER-CONTROL STRUCTURES ASSOCIATED WITH NUCLEAR POWER PLANTS Program Description The RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants Program is an existing program that requires periodic monitoring of water-control structures so that the consequences of age-related deterioration and degradation can be prevented or mitigated in a timely manner. The program contains guidance on engineering data compilation, inspection activities, technical evaluation, inspection frequency, and the content of inspection reports.
NUREG-1801 Consistency The RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants Program, with enhancements, is consistent with the program described in NUREG-1801,Section XI.S7, RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants.
Exceptions to NUREG-1801 None Enhancements The following enhancements will be implemented prior to the period of extended operation.
Elements Affected Enhancements
- 4. Detection of Aging Effects The RG 1.127 Program will be enhanced to clarify that detection of aging effects will monitor accessible structures on a frequency not to exceed five years, consistent with the frequency for implementing the requirements of RG 1.127.
- 4. Detection of Aging Effects The program will be enhanced to perform periodic sampling, testing, and analysis of ground water chemistry for pH, chlorides, and sulfates on a frequency of at least every five years.
- 6. Acceptance Criteria The program will be enhanced to include quantitative acceptance criteria for evaluation and acceptance based on the guidance provided in ACI 349.3R.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-106 Operating Experience A small piece of concrete on the edge of the concrete slab that surrounds the standby service water (SSW) ultimate heat sink basin was discovered to be missing in 2004. An engineering walk-down of the entire perimeter of the basin confirmed this to be the only new piece of missing concrete. The "break area" was found to be relatively clean, with no exposed re-bar or large aggregate. This was evaluated and found to be acceptable with no further action required.
Identification of degradation and evaluation of impact prior to loss of intended function provide evidence that the program is effective for managing aging effects.
Identification of signs of possible degradation and corrective action prior to loss of intended function provides evidence that the program is effective for managing aging effects for components.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants Program has been effective at managing aging effects. The RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-107 B.1.40 SELECTIVE LEACHING Program Description The Selective Leaching Program is a new program that includes a one-time visual inspection of selected components coupled with hardness measurement or other mechanical examination techniques to determine whether loss of material is occurring due to selective leaching.
This inspection will be performed within the five years prior to the period of extended operation.
NUREG-1801 Consistency The Selective Leaching Program will be consistent with the program described in NUREG-1801,Section XI.M33, Selective Leaching of Materials.
Exceptions to NUREG-1801 None Enhancements None Operating Experience The Selective Leaching Program is a new program. Industry operating experience will be considered when implementing this program. Plant operating experience for this program will be gained as it is implemented during the period of extended operations, and will be factored into the program via the confirmation and corrective action elements of the GGNS 10 CFR 50 Appendix B quality assurance program.
This inspection program applies to potential aging effects for which there is currently no operating experience at GGNS indicating the need for an aging management program. As stated in NUREG-1801,Section XI.M33, the inspection elements of this program (e.g., the scope of the inspections and inspection techniques) are consistent with industry practice and will be effective in managing the aging effect included in this program. Accordingly, there is reasonable assurance that this new aging management program will be effective during the period of extended operation.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-108 Conclusion The Selective Leaching Program will be effective for managing aging effects since it will incorporate proven monitoring techniques, acceptance criteria, corrective actions, and administrative controls. The Selective Leaching Program provides assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-109 B.1.41 SERVICE WATER INTEGRITY Program Description The Service Water Integrity Program is an existing program that manages loss of material and fouling in open-cycle cooling water systems as described in the GGNS response to NRC GL 89-
- 13.
NUREG-1801 Consistency The Service Water Integrity Program is consistent with the program described in NUREG-1801,Section XI.M20, Open-Cycle Cooling Water System.
Exceptions to NUREG-1801 None Enhancements None Operating Experience A snapshot program self-assessment in 2002 concluded that the standby service water (SSW) system meets its design and licensing bases and that it is capable of performing its safety functions.
Results of a QA audit in 2003 indicated that the service water integrity program was effective in meeting regulatory requirements and applicable codes and standards.
A program assessment in 2003 identified degraded areas of coating based on inspections of submerged piping in SSW basins. A corrective action plan was initiated and completed to address this issue. Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing aging effects.
In 2004 a strategic plan was established to define monitoring practices and establish a chemical treatment program for the continued improvement in the performance of the SSW system. The plan includes controls for biological fouling, scale formation, solids deposition, and metal corrosion inhibition. Identification of program deficiencies, and subsequent corrective actions, provide assurance that the program will remain effective for managing loss of material of components.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-110 A QA audit in 2007 focused on reviewing responses and action plans to address program issues that had been identified by NRC and by an Entergy corporate assessment the previous year.
Corrective action plans were reviewed and found to contain well-documented causes with timely action plans. This audit confirmed that the service water integrity program was being implemented in a manner that resulted in effective monitoring, inspection, and detection of degradation.
A program assessment in 2009 evaluated the health of the system and the program, corrective action resolution and timeliness, preventive maintenance backlog, leaks and leak repairs, trending and monitoring practices, long-range plans, and operating experience reviews. The report concluded that performance and regulatory margin of the SSW system will be restored once appropriate corrective actions are implemented for certain structural and operational issues. Corrective actions were set forth to address these issues. The execution of aggressive preventive maintenance, inspections and effective chemical treatment assure the long-term integrity of the system.
A QA audit in 2009 confirmed that the service water integrity program was being implemented in a manner that resulted in effective monitoring, inspection, and detection of degradation.
During a visual inspection in 2010, excessive pitting and corrosion was detected on a standby service water valve body and discharge flange. Ultrasonic (UT) examination was performed, and UT data indicated the remaining wall thickness for the valve body was in excess of the minimum wall thicknesses required by ASME Code. Identification and evaluation of aging effects prior to loss of intended function provides evidence that the program remains effective.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The Service Water Integrity Program has been effective at managing aging effects. The Service Water Integrity Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-111 B.1.42 STRUCTURES MONITORING Program Description The Structures Monitoring Program is an existing program that manages the effects of aging on structures and structural components, including structural bolting, within the scope of license renewal. The program was developed based on guidance in Regulatory Guide 1.160 Revision 2, "Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," and NUMARC 93-01 Revision 2, "Industry Guidelines for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," to satisfy the requirement of 10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants."
NUREG-1801 Consistency The Structures Monitoring Program, with enhancements, is consistent with the program described in NUREG-1801,Section XI.S6, Structures Monitoring Program.
Exceptions to NUREG-1801 None Enhancements The following enhancements will be implemented prior to the period of extended operation.
Elements Affected Enhancements
- 1. Scope of Program The Structures Monitoring Program will be enhanced to clarify that the scope will (a) Include the following in-scope structures and structural components.
- Containment building (GGN2)
- Control house - switchyard
- Culvert No. 1 and drainage channel
- Manholes and duct banks
- Radioactive waste building pipe tunnel
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-112 (cont.)
(b) Include the following in-scope structural components.
- Anchor bolts
- Anchorage / embedments
- Base plates
- Basin debris screen and grating
- Battery racks
- Beams, columns, floor slabs and interior walls
- Cable tray and cable tray supports
- Component and piping supports
- Conduit and conduit supports
- Containment sump liner and penetrations
- Containment sump structures
- Control room ceiling support system
- Cooling tower drift eliminators
- Cooling tower fill
- CST/RWST retaining basin (wall)
- Diesel fuel tank access tunnel slab
- Drainage channel
- Drywell floor slab (concrete)
- Drywell wall (concrete)
- Duct banks
- Electrical and instrument panels and enclosures
- Equipment pads/foundations
- Exterior walls
- Fan stack grating
- Fire proofing
- Flood curbs
- Flood retention materials (spare parts)
- Flood, pressure and specialty doors
- Floor slab
- Foundations
- HVAC duct supports
- Instrument line supports
- Instrument racks, frames and tubing trays
- Interior walls
- Main steam pipe tunnel
- Manholes Elements Affected Enhancements
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-113 (cont.)
- Manways, hatches, manhole covers, and hatch covers
- Metal siding
- Missile shields
- Monorails
- Penetration sealant (flood, radiation)
- Penetration sleeves (mechanical/ electrical not penetrating primary containment boundary)
- Pipe whip restraints
- Pressure relief panels
- Reactor pedestal
- Reactor shield wall (steel portion)
- Roof decking
- Roof hatches
- Roof membrane
- Roof slabs
- RPV pedestal sump liner and penetrations
- Seals and gaskets (doors, manways and hatches)
- Seismic isolation joint
- Stairway, handrail, platform, grating, decking, and ladders
- Structural bolting
- Structural steel beams, columns, and plates
- Support members: welds, bolted connections, support anchorages to building structure
- Support pedestals
- Transmission towers1
- Upper containment pool floor and walls
- Vents and louvers (c) Clarify the term significant degradation to include the phrase, that could lead to loss of structural integrity.
(d)Include guidance to perform periodic sampling and analysis of ground water chemistry for pH, chlorides, and sulfates on a frequency of at least once every five years.
Elements Affected Enhancements
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-114
- 3. Parameters Monitored or Inspected The Structures Monitoring Program will be enhanced to clarify that parameters monitored or inspected will (a) Include the inspection for missing nuts for the structural connections.
(b) Include monitoring of sliding/bearing surfaces, such as lubrite plates for loss of material due to wear or corrosion, debris, or dirt. The program will be enhanced to include monitoring elastomeric vibration isolators and structural sealants for cracking, loss of material, and hardening.
- 4. Detection of Aging Effects The Structures Monitoring Program will be enhanced to clarify that detection of aging effects will (a)Include inspection requirements for vibration isolators will be enhanced to include augmented inspections by feel or touch to detect hardening if the vibration isolation function is suspect.
(b)Require inspections every five years for structures and structural components within the scope of license renewal unless technical justification is provided to extend the inspection to a period not to exceed ten years.
- 6. Acceptance Criteria The Structures Monitoring Program acceptance criteria will be enhanced by prescribing acceptance criteria based on information provided in industry codes, standards, and guidelines including NEI 96-03, ACI 201.1R-92, ANSI/ASCE 11-99, and ACI 349.3R-96.
Industry and plant-specific operating experience will also be considered in the development of the acceptance criteria.
- 1. The inspections of these structures may be performed by the transmission personnel. However, the results of the inspections will be provided to the GGNS Structures Monitoring Program owner for review.
Elements Affected Enhancements
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-115 Operating Experience During a 2007 inspection, Door 1M112 at the standby service water valve room would not open completely due to sinking of the concrete enclosure over the door. A review of reports from 1990 and 1998 indicated a gradual vertical settlement in this area. A review of the applicable design calculation indicated that the vertical settlement was not an issue for the shield wall, although it could interfere with the ability to use the door. An engineering change was prepared to allow removal of concrete in the area where the door is rubbing. Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing aging effects for structural components.
Water leaking in the reactor water cleanup (RWCU) heat exchanger room from a crack in the ceiling was identified during an RF15 walkdown in 2007. This crack had been earlier identified as a result of shrinkage in 1987. Samples of the leakage were tested for iron with negative results, indicating that corrosion of the concrete reinforcing steel was not occurring. The reinforced concrete structure of the RWCU heat exchanger room was found to be structurally adequate. Evaluation of degradation prior to loss of intended function provides evidence that the program is effective for managing aging effects for structural components.
Structures Monitoring periodic assessments covering the period from 2001 through 2007 identified no major structural problems. Corrective actions were set forth for minor structural degradations. The assessments concluded the Structures Monitoring Program is adequate and effective.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The Structures Monitoring Program has been effective at managing aging effects. The Structures Monitoring Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-116 B.1.43 WATER CHEMISTRY CONTROL - BWR Program Description The Water Chemistry Control - BWR Program is an existing program that manages loss of material, cracking, and fouling in components exposed to a treated water environment through monitoring and control of water chemistry. EPRI water chemistry guidelines are used.
The One-Time Inspection Program utilizes inspections or non-destructive evaluations of representative samples to verify that the Water Chemistry Control - BWR Program has been effective at managing aging effects.
NUREG-1801 Consistency The Water Chemistry Control - BWR Program is consistent with the program described in NUREG-1801,Section XI.M2, Water Chemistry.
Exceptions to NUREG-1801 None Enhancements None Operating Experience SOER 03-02, Managing Core Design Changes, was issued to address unsuccessful industry efforts to obtain defect-free fuel performance. This SOER required an evaluation of the effects of chemistry changes on core and fuel performance and the effects of core and fuel design changes on coolant chemistry. This evaluation required a review of chemistry-related issues and how these issues are addressed in the Water Chemistry Control - BWR Program. The results of this review and the required responses upgraded and confirmed the effectiveness of the program.
Identification of program deficiencies, and subsequent corrective actions, provide assurance that the program will remain effective for managing loss of material of components.
In 2006 a QA audit indicated that the chemistry program was effective in meeting intended results. The program was found to adequately prevent chemistry excursions related to reactor water sulfates and feedwater iron and to provide protection of plant components through effective chemistry control, monitoring, calculation, and reporting of chemistry data.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-117 During the period from 2007 through 2010, several condition reports were initiated due to adverse trends in parameters monitored by the Water Chemistry Control - BWR Program.
Corrective actions were taken within the Corrective Action Program to preclude reaching unacceptable values for the parameters monitored. No impact on plant materials was experienced. The routine confirmation of water quality and use of appropriate timely corrective action provide evidence that the program is effective in managing loss of material for applicable components.
An assessment of the effectiveness of corrective actions for elevated sulfate concentration in the reactor coolant was performed in 2008. This included a review of analytical data, post-corrective-action baseline sulfate concentration, and comparison of analytical data for sulfate performance during period 2007-2008 during steady state, transient state and hot weather conditions. The assessment concluded the actions taken to reduce the level of sulfate concentration in the reactor coolant were successful and included new actions and recommendations which were resolved to upgrade the program to enhance its effectiveness.
In 2009 a strategic plan was established to implement water chemistry initiatives. The plan optimized corrosion control for the reactor vessel, primary system components, and balance of plant (BOP) components. It was based upon industry experience and guidelines, BWR cycle design, and BWR metallurgy. The plan emphasized the reduction of intergranular stress corrosion cracking in primary system components, minimization of flow accelerated corrosion in BOP systems, high standards of fuel integrity, and minimization of radiation field buildup. An advanced resin cleaning system was implemented to improve the cleaning of condensate resins.
Condensate temperatures were managed in cold weather months to improve iron removal.
Identification of needed program enhancements, and subsequent corrective actions, provide assurance that the program will remain effective for managing loss of material, cracking, and fouling of components.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The Water Chemistry Control - BWR Program has been effective at managing aging effects. The Water Chemistry Control - BWR Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-118 B.1.44 WATER CHEMISTRY CONTROL - CLOSED TREATED WATER SYSTEMS Program Description The Water Chemistry Control - Closed Treated Water Systems Program is an existing program that manages loss of material, cracking, and fouling in components exposed to a treated water environment through monitoring and control of water chemistry as well as visual inspections.
NUREG-1801 Consistency The Water Chemistry Control - Closed Treated Water Systems Program, with enhancements, is consistent with the program described in NUREG-1801,Section XI.M21A, Closed Treated Water Systems.
Exceptions to NUREG-1801 None Enhancements The following enhancements will be implemented prior to the period of extended operation.
Elements Affected Enhancements
- 1. Scope of Program
- 2. Preventive Actions The Water Chemistry Control - Closed Treated Water Systems Program will be enhanced to provide a corrosion inhibitor for the engine jacket water on the engine-driven fire water pump diesels in accordance with industry guidelines and vendor recommendations.
- 3. Parameters Monitored or Inspected The Water Chemistry Control - Closed Treated Water Systems Program will be enhanced to provide periodic flushing of the engine jacket water and cleaning of heat exchanger tubes for the engine-driven fire water pump diesels in accordance with industry guidelines and vendor recommendations.
- 4. Detection of Aging Effects The Water Chemistry Control - Closed Treated Water Systems Program will be enhanced to provide testing of the engine jacket water for the engine-driven fire water pump diesels at least once per refueling cycle.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-119
- 4. Detection of Aging Effects The Water Chemistry Control - Closed Treated Water Systems Program will be enhanced to conduct inspections whenever a boundary is opened for the following systems.
- Drywell chilled water (DCW, system P72)
- Plant chilled water (PCW, system P71)
- Diesel generator cooling water subsystem for Division I and II standby diesel generators
- Diesel engine jacket water for engine-driven fire water pumps
- Diesel generator cooling water subsystem for Division III (HPCS) diesel generator
- Turbine building cooling water (TBCW, system P43)
- Component cooling water (CCW, system P42)
These inspections will be conducted in accordance with applicable ASME Code requirements, industry standards, and other plant-specific inspection and personnel qualification procedures that are capable of detecting corrosion or cracking.
Elements Affected Enhancements
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-120 Operating Experience In 2007 a strategic plan was established to define monitoring practices and a chemical treatment program for the continued improvement in the performance of plant systems in the Water Chemistry Control - Closed Treated Water Systems Program. The plan set forth a program that will minimize corrosion, cost, monitoring requirements and frequent adjustments.
- 4. Detection of Aging Effects The Water Chemistry Control -- Closed Treated Water Systems Program will be enhanced to inspect a representative sample of piping and components at a frequency of once every ten years for the following systems.
- Drywell chilled water (DCW, system P72)
- Plant chilled water (PCW, system P71)
- Diesel generator cooling water subsystem for Division I and II standby diesel generators
- Diesel engine jacket water for engine-driven fire water pumps
- Diesel generator cooling water subsystem for Division III (HPCS) diesel generator
- Turbine building cooling water (TBCW, system P43)
- Component cooling water (CCW, system P42)
Components inspected will be those with the highest likelihood of corrosion or cracking. A representative sample is 20% of the population (defined as components having the same material, environment, and aging effect combination) with a maximum of 25 components. The inspection methods will be in accordance with applicable ASME Code requirements, industry standards, or other plant-specific inspection and personnel qualification procedures that ensure the capability of detecting corrosion or cracking.
Elements Affected Enhancements
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-121 During the period from 2008 through 2010, several condition reports were initiated due to adverse trends in parameters monitored by the Water Chemistry Control - Closed Treated Water Systems Program. Corrective actions were taken within the Corrective Action Program to preclude reaching unacceptable values for the parameters monitored. The routine confirmation of water quality and use of appropriate timely corrective action provide evidence that the program is effective in managing loss of material for applicable components.
The process for review of future plant-specific and industry operating experience for this program is discussed in Section B.0.4.
Conclusion The Water Chemistry Control - Closed Treated Water Systems Program has been effective at managing aging effects. The Water Chemistry Control - Closed Treated Water Systems Program assures the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.
Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix B Aging Management Programs and Activities Page B-122 B.2 REFERENCES B.2-1 U.S. Nuclear Regulatory Commission, NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, Revision 2, December 2010.
B.2-2 U.S. Nuclear Regulatory Commission, NUREG-1801, Generic Aging Lessons Learned (GALL) Report, Revision 2, December 2010.