ML11308A095

License Renewal Application Technical Information - Appendix a - Updated Final Safety Analysis Report Supplement

ML11308A095
Person / Time
Site:	Grand Gulf
Issue date:	10/28/2011
From:	Entergy Operations
To:	Office of Nuclear Reactor Regulation
References
GNRO-2011/00093
Download: ML11308A095 (37)

Text

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-i APPENDIX A UPDATED FINAL SAFETY ANALYSIS REPORT SUPPLEMENT TABLE OF CONTENTS A.0 INTRODUCTION.................................................... A-1 A.1 AGING MANAGEMENT PROGRAMS.................................... A-1 A.1.1 115 kV Inaccessible Transmission Cable Program...................... A-2 A.1.2 Aboveground Metallic Tanks Program................................ A-2 A.1.3 Bolting Integrity Program.......................................... A-3 A.1.4 Boraflex Monitoring Program....................................... A-3 A.1.5 Buried Piping and Tanks Inspection Program.......................... A-4 A.1.6 BWR CRD Return Line Nozzle Program.............................. A-4 A.1.7 BWR Feedwater Nozzle Program................................... A-4 A.1.8 BWR Penetrations Program........................................ A-4 A.1.9 BWR Stress Corrosion Cracking Program............................. A-4 A.1.10 BWR Vessel ID Attachment Welds Program........................... A-4 A.1.11 BWR Vessel Internals Program..................................... A-5 A.1.12 Compressed Air Monitoring Program................................. A-5 A.1.13 Containment Inservice Inspection - IWE Program...................... A-6 A.1.14 Containment Inservice Inspection - IWL Program....................... A-6 A.1.15 Containment Leak Rate Program.................................... A-6 A.1.16 Diesel Fuel Monitoring Program..................................... A-7 A.1.17 Environmental Qualification (EQ) of Electric Components Program......... A-7 A.1.18 External Surfaces Monitoring Program............................... A-8 A.1.19 Fatigue Monitoring Program........................................ A-8 A.1.20 Fire Protection Program........................................... A-9 A.1.21 Fire Water System Program........................................ A-9 A.1.22 Flow-Accelerated Corrosion Program................................ A-10 A.1.23 Inservice Inspection Program....................................... A-11 A.1.24 Inservice Inspection - IWF Program................................. A-11

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-ii A.1.25 Inspection of Overhead Heavy Load and Light Load (Related to Refueling)

Handling Systems Program........................................ A-12 A.1.26 Internal Surfaces in Miscellaneous Piping and Ducting Components Program. A-12 A.1.27 Masonry Wall Program............................................ A-13 A.1.28 Non-EQ Cable Connections Program................................ A-13 A.1.29 Non-EQ Inaccessible Power Cables (400 V to 35 kV) Program............ A-14 A.1.30 Non-EQ Instrumentation Circuits Test Review Program.................. A-14 A.1.31 Non-EQ Insulated Cables and Connections Program.................... A-15 A.1.32 Oil Analysis Program............................................. A-15 A.1.33 One-Time Inspection Program...................................... A-16 A.1.34 One-Time Inspection - Small-Bore Piping Program..................... A-17 A.1.35 Periodic Surveillance and Preventive Maintenance Program.............. A-18 A.1.36 Protective Coating Monitoring and Maintenance Program................. A-18 A.1.37 Reactor Head Closure Studs Program................................ A-19 A.1.38 Reactor Vessel Surveillance Program................................ A-19 A.1.39 RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants Program............................................ A-19 A.1.40 Selective Leaching Program....................................... A-20 A.1.41 Service Water Integrity Program.................................... A-20 A.1.42 Structures Monitoring Program..................................... A-20 A.1.43 Water Chemistry Control - BWR Program............................. A-23 A.1.44 Water Chemistry Control - Closed Treated Water Systems Program........ A-23 A.2 EVALUATION OF TIME-LIMITED AGING ANALYSES....................... A-25 A.2.1 Reactor Vessel Neutron Embrittlement............................... A-25 A.2.1.1 Reactor Vessel Fluence................................... A-25 A.2.1.2 Pressure-Temperature Limits............................... A-25 A.2.1.3 Upper-Shelf Energy....................................... A-25 A.2.1.4 Reactor Vessel Circumferential Weld Inspection Relief........... A-26 A.2.1.5 Reactor Vessel Axial Weld Failure Probability.................. A-26 A.2.1.6 Reactor Pressure Vessel Core Reflood Thermal Shock Analysis.... A-26 A.2.2 Metal Fatigue................................................... A-27 A.2.2.1 Class 1 Metal Fatigue..................................... A-27

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-iii A.2.2.2 Non-Class 1 Metal Fatigue................................. A-29 A.2.2.3 Effects of Reactor Water Environment on Fatigue Life............ A-30 A.2.3 Environmental Qualification of Electrical Components................... A-31 A.2.4 Fatigue of Primary Containment, Attached Piping, and Components........ A-31 A.2.5 Other Plant-Specific TLAA......................................... A-32 A.2.5.1 Erosion of the Main Steam Line Flow Restrictors................ A-32 A.2.5.2 Determination of Intermediate High-Energy Line Break Locations... A-33 A.2.5.3 Fluence Effects for the Reactor Vessel Internals................ A-33 A.3 REFERENCES...................................................... A-34

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-1 A.0 INTRODUCTION This appendix provides the information to be submitted in an Updated Final Safety Analysis Report (UFSAR) Supplement as required by 10 CFR 54.21(d) for the Grand Gulf Nuclear Station (GGNS) License Renewal Application (LRA). Appendix B of the GGNS LRA provides descriptions of the programs and activities that manage the effects of aging for the period of extended operation. Section 4 of the LRA documents the evaluations of time-limited aging analyses for the period of extended operation. Appendix B and Section 4 have been used to prepare the summary program and activity descriptions for this appendix.

The information presented in this section will be incorporated into the UFSAR following issuance of the renewed operating license. Upon inclusion of the UFSAR Supplement in the GGNS UFSAR, future changes to the descriptions of the programs and activities will be made in accordance with 10 CFR 50.59.

The following information documents aging management programs and activities credited in the Grand Gulf Nuclear Station (GGNS) license renewal review (Section A.1) and time-limited aging analyses evaluated for the period of extended operation (Section A.2).

A AGING MANAGEMENT PROGRAMS AND ACTIVITIES The GGNS license renewal application (Reference A.3-1) and information in subsequent related correspondence provided sufficient basis for the NRC to make the findings required by 10 CFR 54.29 (Final Safety Evaluation Report) (Reference A.3-2). As required by 10 CFR 54.21(d), this UFSAR supplement contains a summary description of the programs and activities for managing the effects of aging (Section A.1) and a description of the evaluation of time-limited aging analyses for the period of extended operation (Section A.2). The period of extended operation is the 20 years after the expiration date of the original operating license.

A.1 AGING MANAGEMENT PROGRAMS The integrated plant assessment for license renewal identified aging management programs necessary to provide reasonable assurance that components within the scope of license renewal will continue to perform their intended functions consistent with the current licensing basis (CLB) for the period of extended operation. This section describes the aging management programs and activities required during the period of extended operation. Aging management programs will be implemented prior to entering the period of extended operation.

GGNS quality assurance (QA) procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR 50, Appendix B. The GGNS Quality Assurance Program applies to 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 and Document Control Program and are applicable to all aging

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-2 management programs and activities during the period of extended operation. The confirmation process is part of the Corrective Action Program and includes reviews to assure adequacy of corrective actions, tracking and reporting of open corrective actions, and review of corrective action effectiveness. Any follow-up inspection required by the confirmation process is documented in accordance with the Corrective Action Program. The corrective action, confirmation process, and administrative 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.

The Operating Experience Program (OEP) and the Corrective Action Program (CAP) help to assure continued effectiveness of aging management programs through evaluations of operating experience. The OEP implements the requirements of NRC NUREG-0737, Clarification of TMI Action Plan Requirements, Section l.C.5, and evaluates site, Entergy fleet, and industry operating experience for impact on GGNS. The CAP implements the requirements of 10 CFR 50, Appendix B, Criterion XVI and is used to evaluate and effect appropriate actions in response to operating experience relevant to GGNS that indicates a condition adverse to quality or a non-conformance.

A.1.1 115 kV Inaccessible Transmission Cable Program The 115 kV Inaccessible Transmission Cable Program manages the effects of aging on the 115 kV inaccessible transmission cable systems. The program includes periodic actions to prevent inaccessible transmission cables from being exposed to significant moisture. In this program, inaccessible 115 kV transmission 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 may be adjusted based on test results and operating experience. The specific type of test will be 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 inspection 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.

This program will be implemented prior to the period of extended operation. The first cable tests and manhole inspections are to be completed prior to the period of extended operation.

A.1.2 Aboveground Metallic Tanks Program The Aboveground Metallic Tanks Program manages 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.

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-3 A.1.3 Bolting Integrity Program The Bolting Integrity Program 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.

The Bolting Integrity Program will be enhanced as follows.

Clarify prohibition on use of lubricants containing molybdenum disulfide (MoS2) for bolting and specify that proper gasket compression will be visually verified following assembly.

The scope of this enhancement will include applicable GGNS site procedures.

Include consideration of the guidance applicable for pressure boundary bolting in NUREG-1339, EPRI NP-5769, and EPRI TR-104213.

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.

Include guidance from EPRI NP-5769 and EPRI TR-104213 for replacement of bolting.

Enhancements will be implemented prior to the period of extended operation.

A.1.4 Boraflex Monitoring Program The Boraflex Monitoring Program manages the change in material properties (neutron-absorbing capacity) in the Boraflex material affixed to spent fuel racks using silica sampling, areal testing activities, 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.

The Boraflex Monitoring Program will be enhanced as follows.

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

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-4 ensure current Region I storage locations will not need to be reclassified as Region II storage locations in the analysis interval.

Enhancements will be implemented prior to the period of extended operation.

A.1.5 Buried Piping and Tanks Inspection Program The Buried Piping and Tanks Inspection Program 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.

A.1.6 BWR CRD Return Line Nozzle Program The BWR Control Rod Drive (CRD) Return Line Nozzle Program manages cracking on the intended function 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.

A.1.7 BWR Feedwater Nozzle Program The BWR Feedwater Nozzle Program 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.

A.1.8 BWR Penetrations Program The BWR Penetrations Program 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.

A.1.9 BWR Stress Corrosion Cracking Program The BWR Stress Corrosion Cracking Program 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.

A.1.10 BWR Vessel ID Attachment Welds Program The BWR Vessel ID Attachment Welds Program 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.

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-5 A.1.11 BWR Vessel Internals Program The BWR Vessel Internals Program 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.

The BWR Vessel Internals Program will be enhanced as follows.

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.

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.

Enhancements will be implemented prior to the period of extended operation.

A.1.12 Compressed Air Monitoring Program The Compressed Air Monitoring Program will be enhanced as follows.

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.

Include periodic and opportunistic inspections of accessible internal surfaces of piping and components in the following compressed air systems.

Automatic Depressurization System (ADS) air Division 1 Diesel Generator Starting Air (D1DGSA)

Division 2 Diesel Generator Starting Air (D2DGSA)

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-6 Division 3 Diesel Generator Starting Air (D3DGSA), also known as the HPCS Diesel Generator Instrument Air (IA) - system P53 Enhancements will be implemented prior to the period of extended operation.

A.1.13 Containment Inservice Inspection - IWE Program The Containment Inservice Inspection - IWE Program is a general visual examination that assesses the condition of the containment steel liner and detects 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.

A.1.14 Containment Inservice Inspection - IWL Program The Containment Inservice Inspection - IWL Program is a general visual examinations that assesses the overall condition of the containment concrete and detects 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.

A.1.15 Containment Leak Rate Program The Containment Leak Rate Program 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

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-7 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.

A.1.16 Diesel Fuel Monitoring Program The Diesel Fuel Monitoring Program 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.

The Diesel Fuel Monitoring Program will be enhanced as follows.

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 10-year period prior to the period of extended operation and at succeeding 10-year intervals. If visual inspection is not possible, a volumetric inspection will be performed.

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 10-year period prior to the period of extended operation and at succeeding 10-year intervals.

Enhancements will be implemented prior to the period of extended operation.

A.1.17 Environmental Qualification (EQ) of Electric Components Program The Environmental Qualification (EQ) of Electric Components 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.

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-8 A.1.18 External Surfaces Monitoring Program The External Surfaces Monitoring Program 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.

The External Surfaces Monitoring Program will be enhanced as follows.

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.

Clearly identify underground components within the scope of this program in program documents. Underground components are those for which access is physically restricted.

Provide instructions for inspecting all underground components within the scope of this program during each 5-year period, beginning 10 years prior to the entry into the period of extended operation.

Enhancements will be implemented prior to the period of extended operation.

A.1.19 Fatigue Monitoring Program The Fatigue Monitoring Program 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.

The Fatigue Monitoring Program will be enhanced as follows.

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.

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 A.2.2.3.

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-9 Program guidance documents will be revised 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.

Enhancements will be implemented at least two years prior to entering the period of extended operation.

A.1.20 Fire Protection Program The Fire Protection Program 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.

The Fire Protection Program will be enhanced as follows.

Require visual inspections of the Halon/CO2 fire suppression system at least once every fuel cycle to examine for signs of corrosion.

Require visual inspections of fire damper framing at least once every fuel cycle to check for signs of degradation.

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.

Enhancements will be implemented prior to the period of extended operation.

A.1.21 Fire Water System Program The Fire Water System Program manages loss of material and fouling for components in fire protection systems using preventive, inspection, and monitoring activities, including periodic full-flow flush test and testing or replacement of sprinkler heads. Applicable industry standards and guidance documents are used to delineate the program.

The Fire Water System Program will be enhanced as follows.

Include periodic visual inspection of spray and sprinkler system internals for evidence of degradation. Acceptance criteria will be enhanced to verify no unacceptable degradation.

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 A Updated Safety Analysis Report Supplement Page A-10 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.

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 10 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.

A representative sample of sprinkler heads will be tested or replaced 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.

Enhancements will be implemented prior to the period of extended operation.

A.1.22 Flow-Accelerated Corrosion Program The Flow-Accelerated Corrosion (FAC) Program 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.

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-11 The FAC Program will be enhanced as follows.

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 chromium material.

This enhancement will be implemented prior to the period of extended operation.

A.1.23 Inservice Inspection Program The Inservice Inspection Program 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.

A.1.24 Inservice Inspection - IWF Program The Inservice Inspection - IWF Program manages aging effects for ASME Class 1, 2, 3 piping and component supports. The scope of inspection for component supports is based on sampling of piping supports and 100 percent of component supports other than piping as specified in Table IWF-2500-1.

The Inservice Inspection - IWF Program will be enhanced as follows.

Address inspections of accessible sliding surfaces.

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. 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.

Clarify that detection of aging will include:

a) Monitoring 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.

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-12 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 Include the following as unacceptable conditions.

a) Loss of material due to corrosion or wear, which reduces the load bearing capacity of the component support.

b) Debris, dirt, or excessive wear that could prevent or restrict sliding of the sliding surfaces as intended in the design basis of the support.

c)

Cracked or sheared bolts, including high strength bolts, and anchors.

Enhancements will be implemented prior to the period of extended operation.

A.1.25 Inspection of Overhead Heavy Load and Light Load (Related to Refueling)

Handling Systems Program The Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems Program consists of periodic inspections and preventive maintenance to manage loss of material for 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.

The Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems Program will be enhanced as follows.

The scope will include monitoring of rails in the rail system for the aging effect wear and structural connections/bolting for loose or missing bolts, nuts, pins or rivets. Additionally, 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.

Revise acceptance criteria 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.

Enhancements will be implemented prior to the period of extended operation.

A.1.26 Internal Surfaces in Miscellaneous Piping and Ducting Components Program The Internal Surfaces in Miscellaneous Piping and Ducting Components Program manages the effects of aging using visual inspections of the internal surfaces of piping and components during

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-13 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.

A.1.27 Masonry Wall Program The Masonry Wall Program 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.

The Masonry Wall Program will be enhanced as follows.

Monitor gaps between the supports and masonry walls that could potentially affect wall qualification.

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.

Enhancements will be implemented prior to the period of extended operation.

A.1.28 Non-EQ Cable Connections Program The Non-EQ Cable Connections Program is a 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 quantitative 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.

This program will be completed prior to the period of extended operation.

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-14 A.1.29 Non-EQ Inaccessible Power Cables (400 V to 35 kV) Program The Non-EQ Inaccessible Power Cables (400 V to 35kV) Program manages the aging effects on the inaccessible power (400 V to 35kV) cable systems. The 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 are 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 inspection 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.

The Non-EQ Inaccessible Power Cables (400 V to 35 kV) Program will be enhanced as follows.

Include low-voltage (400 V to 2 kV) power cables.

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.

Clarify 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.

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.

A.1.30 Non-EQ Instrumentation Circuits Test Review Program The Non-EQ Instrumentation Circuits Test Review Program manages the aging effects of the applicable cables in the neutron monitoring and process radiation monitoring systems or sub-systems. The 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.

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-15 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.

A.1.31 Non-EQ Insulated Cables and Connections Program The Non-EQ Insulated Cables and Connections Program 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, melting, swelling, or surface contamination. The program sample consists of all accessible cables and connections in localized adverse environments, and 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 ten years, with the first inspection prior to the period of extended operation.

This program will be implemented prior to the period of extended operation.

A.1.32 Oil Analysis Program The Oil Analysis Program 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.

The Oil Analysis Program will be enhanced as follows.

Include piping and components within the main generator system (N41) with an internal environment of lube oil.

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-16 Provide a formalized analysis technique for particulate counting.

Enhancements will be implemented prior to the period of extended operation.

A.1.33 One-Time Inspection Program The One-Time Inspection Program 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.

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-17 The inspection will be performed within the ten years prior to the period of extended operation.

A.1.34 One-Time Inspection - Small-Bore Piping Program The One-Time Inspection - Small-Bore Piping Program 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. 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.

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 A Updated Safety Analysis Report Supplement Page A-18 A.1.35 Periodic Surveillance and Preventive Maintenance Program The Periodic Surveillance and Preventive Maintenance Program 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.

Gasket/seal for upper containment pool gates in containment building.

Low pressure core spray system (LPCS) piping passing through the waterline region of suppression pool.

Residual heat removal (RHR) system piping passing through the waterline region of suppression pool.

Pressure relief system piping passing through the waterline region of the suppression pool.

Reactor core isolation cooling (RCIC) system piping passing through the waterline region of the suppression pool.

Control rod drive (CRD) system piping.

Circulating water system piping and valve bodies.

Floor and equipment drain system piping, drain housings, and valve bodies.

High pressure core spray (HPCS) system piping passing through the waterline region of the suppression pool.

Floor and equipment drain system piping below the waterline in the in-scope sumps.

The Periodic Surveillance and Preventive Maintenance Program will be enhanced as follows.

Revise program guidance documents as necessary to assure that the 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.

This enhancement will be implemented prior to the period of extended operation.

A.1.36 Protective Coating Monitoring and Maintenance Program The Protective Coating Monitoring and Maintenance Program monitors and maintains service level I coatings inside containment. The program assesses coating condition through visual inspections.

The Protective Coating Monitoring and Maintenance Program will be enhanced as follows.

Revise program documents to include parameters monitored or inspected per the guidance provided in ASTM D5163-08.

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-19 Revise program documents to provide for inspection of coatings near sumps or screens associated with the emergency core cooling system.

Revise program documents to include acceptance criteria per ASTM D 5163-08.

Enhancements will be implemented prior to the period of extended operation.

A.1.37 Reactor Head Closure Studs Program The Reactor Head Closure Studs Program 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.

A.1.38 Reactor Vessel Surveillance Program The Reactor Vessel Surveillance Program 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.

The Reactor Vessel Surveillance Maintenance Program will be enhanced as follows.

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.

This enhancement will be implemented prior to the period of extended operation.

A.1.39 RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants Program 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.

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-20 The RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants Program will be enhanced as follows.

Accessible structures will be monitored on a frequency not to exceed five years, consistent with the frequency for implementing the requirements of RG 1.127.

Perform periodic sampling, testing, and analysis of ground water chemistry for pH, chlorides, and sulfates on a frequency of at least every five years.

Include quantitative acceptance criteria for evaluation and acceptance based on the guidance provided in ACI 349.3R.

Enhancements will be implemented prior to the period of extended operation.

A.1.40 Selective Leaching Program The Selective Leaching Program 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.

A.1.41 Service Water Integrity Program The Service Water Integrity Program manages loss of material and fouling in open-cycle cooling water systems as described in the GGNS response to NRC GL 89-13.

A.1.42 Structures Monitoring Program The Structures Monitoring Program 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."

The Structures Monitoring Program will be enhanced as follows.

Clarify that the scope includes the following in-scope structures and structural components.

Containment Building (GGN 2)

Control House - Switchyard

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-21 Culvert No. 1 and drainage channel Manholes and duct banks Radioactive waste building pipe tunnel Clarify that the scope includes 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 Manways, hatches, manhole covers, and hatch covers Metal siding

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-22 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 Sumps and sump liners Support members: welds, bolted connections, support anchorages to building structure Support pedestals Transmission towers (see Note 1)

Upper containment pool floor and walls Vents and louvers Note 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.

Clarify the term "significant degradation" to include the following: "that could lead to loss of structural integrity...."

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.

Include an inspection for missing nuts for the structural connections.

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.

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-23 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.

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.

Prescribe 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.

Enhancements will be implemented prior to the period of extended operation.

A.1.43 Water Chemistry Control - BWR Program The Water Chemistry Control - BWR Program 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.

A.1.44 Water Chemistry Control - Closed Treated Water Systems Program The Water Chemistry Control - Closed Treated Water Systems Program 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.

The Water Chemistry Control - Closed Treated Water Systems Program will be enhanced as follows.

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.

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.

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 A Updated Safety Analysis Report Supplement Page A-24 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.

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.

Enhancements will be implemented prior to the period of extended operation.

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-25 A.2 EVALUATION OF TIME-LIMITED AGING ANALYSES In accordance with 10 CFR 54.21(c), an application for a renewed license requires an evaluation of time-limited aging analyses for the period of extended operation. The following time-limited aging analyses have been identified and evaluated to meet this requirement.

A.2.1 Reactor Vessel Neutron Embrittlement The reactor vessel neutron embrittlement time-limited aging analyses including consideration for extended power uprate (EPU) either have been projected to the end of the period of extended operation in accordance with 10 CFR 54.21(c)(1)(ii) or will be managed for the period of extended operation in accordance with 10 CFR 54.24(c)(1)(iii) as summarized below.

Based on the plant operating history, a projected EFPY value of 54 EFPY is used to evaluate reactor vessel neutron embrittlement time-limited aging analyses.

A.2.1.1 Reactor Vessel Fluence Calculated fluence is based on a time-limited assumption defined by the operating term.

Therefore, analyses that evaluate reactor vessel neutron embrittlement based on calculated fluence are time-limited aging analyses.

The high-energy (> 1 MeV) neutron fluence for the nozzles, welds and shells of the reactor pressure vessel (RPV) beltline region was determined using the General Electric-Hitachi (GEH) method for neutron flux calculation documented in report NEDC-32983P-A and approved by the NRC. The method adheres to the guidance prescribed in Regulatory Guide (RG) 1.190 as was described in the EPU submittal. (Reference A.3-3).

A.2.1.2 Pressure-Temperature Limits Appendix G of 10 CFR 50 requires that the reactor vessel remain within established pressure-temperature (P-T) limits during boltup, hydro-test, pressure tests, normal operation, and anticipated operational occurrences. These limits are calculated using materials and fluence data, including data obtained through the Reactor Vessel Surveillance Program.

The P-T limit curves will continue to be updated, as required by Appendix G of 10 CFR Part 50, assuring that limits remain valid through the period of extended operation.

The time-limited aging analyses for reactor vessel pressure-temperature limits will be managed for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(iii).

A.2.1.3 Upper-Shelf Energy The predictions for percent drop in upper shelf energy (USE) values were projected to 54 EFPY using projected beltline fluence values, chemistry and surveillance data, and un-irradiated USE

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-26 information in accordance with Regulatory Guide 1.99. All projected USE values for 54 EFPY remain above the 50 ft-lb minimum acceptable value specified in Appendix G of 10 CFR 50.

The time-limited aging analyses for upper shelf energy have been projected to the end of the period of extended operation in accordance with 10 CFR 54.21(c)(1)(ii).

A.2.1.4 Reactor Vessel Circumferential Weld Inspection Relief The GGNS reactor pressure vessel circumferential weld parameters at 54 EFPY will remain within the NRC's (64 EFPY) bounding parameters from the BWRVIP-05 SER. The fact that the values projected to the end of the period of extended operation are less than the 64 EFPY value provided by the NRC leads to the conclusion that the GGNS RPV conditional failure probability is less than the conditional failure probability of the NRC analysis. As such, the conditional probability of failure for circumferential welds remains below that determined during the NRC's final safety evaluation of BWRVIP-05.

The reactor vessel circumferential weld inspection relief for the extended operating period will be submitted to the NRC in accordance with 10 CFR 50.55(a).

The time-limited aging analysis for reactor vessel circumferential weld inspection relief has been projected to the end of the period of extended operation in accordance with 10 CFR 54.21(c)(1)(ii).

A.2.1.5 Reactor Vessel Axial Weld Failure Probability The NRC SER for BWRVIP-74-A evaluated the failure frequency of axially oriented welds in BWR reactor vessels. Applicants for license renewal must evaluate axially oriented RPV welds to show that their failure frequency remains below the value calculated in the BWRVIP-74 SER.

The SER states that an acceptable way to do this is to show that the mean RTNDT of the limiting axial beltline weld at the end of the period of extended operation is less than the values specified in the SER.

The projected 54 EFPY GGNS mean ART is less than the bounding value shown in the NRC SER for BWRVIP-74.

Reactor vessel axial weld TLAA has been projected through the period of extended operation in accordance with 10 CFR 54.21(c)(1)(ii).

A.2.1.6 Reactor Pressure Vessel Core Reflood Thermal Shock Analysis General Electric Report NEDO-10029 is referenced in Section 5.3.3 of the UFSAR. NEDO-10029 addressed the concern for brittle fracture of the reactor pressure vessel due to reflood following a postulated loss of coolant accident (LOCA). In addition to the NEDO-10029 that is listed in the UFSAR, there is a more recent analysis of the BWR-6 vessels (Ranganath, S.,

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-27 "Fracture Mechanics Evaluation of a Boiling Water Reactor Vessel Following a Postulated Loss of Coolant Accident," Fifth International Conference on Structural Mechanics in Reactor Technology, Berlin, Germany, August 1979). The more recent analysis is appropriate for the GGNS reactor pressure vessel because it evaluates the bounding LOCA event, a main steam line break, for a BWR-6 vessel design.

This analysis shows that when the peak stress intensity occurs at approximately 300 seconds after the LOCA, the temperature inside the vessel wall is approximately 400ºF. The maximum ART value calculated for the GGNS RPV beltline material is 53ºF. Using the equation for KIC presented in Appendix A of ASME Section XI and the maximum ART value, the material reaches upper shelf at approximately158ºF, which is well below the minimum 400ºF temperature predicted for the thermal shock event at the time of peak stress intensity. Therefore, the revised analysis has projected the TLAA through the period of extended operation. The time-limited aging analysis for Reactor Pressure Vessel Core Reflood Thermal Shock Analysis has been projected to the end of the period of extended operation in accordance with 10 CFR 54.21(c)(1)(ii).

A.2.2 Metal Fatigue A.2.2.1 Class 1 Metal Fatigue Fatigue evaluations were performed in the design of the GGNS Class 1 components in accordance with their design requirements. ASME Section III fatigue evaluations are contained in analyses and stress reports, and because they may be based on a number of transient cycles assumed for a 40-year operating term, these evaluations are considered time-limited aging analyses.

Design cyclic loadings and thermal conditions for the Class 1 components are defined by the applicable design specifications for each component. The original design specifications provided a set of transients that were used in the design of the components and are included as part of each component analysis or stress report.

The Fatigue Monitoring Program tracks and evaluates the cycles and requires corrective actions if limits are approached.

Reactor Vessel As described in UFSAR Section 5.3.3.3, the reactor pressure vessel is a vertical, cylindrical pressure vessel of welded construction fabricated in accordance with ASME Code,Section III, Class 1 requirements. Fatigue evaluations for the reactor vessel were performed as part of the vessel design. The fatigue analyses of the reactor vessel are considered time-limited aging analyses because they are based on numbers of design cycles that were expected to occur in 40 years of operation.

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-28 GGNS will monitor these transient cycles using the Fatigue Monitoring Program and assure that action is taken if any of the actual cycles approach their analyzed numbers. As such, the Fatigue Monitoring Program will manage the effects of aging due to fatigue on the reactor vessel in accordance with 10 CFR 54.21(c)(1)(iii).

Reactor Vessel Feedwater Nozzle As described in UFSAR Section 5.3.3.1.4.5.1, GGNS implemented a plant modification prior to plant operation to eliminate concerns identified in previous BWR designs. A second piston ring and triple thermal sleeves have been incorporated in the design for Grand Gulf.

The analysis of the modified feedwater nozzle included fatigue from potential rapid cycling behind the thermal sleeves. Therefore, for the FW nozzle there is a location-specific rapid cycling fatigue usage that added to the cycle-based fatigue. The usage is postulated based on time and feedwater temperature in order to include the rapid cycling effect.

The feedwater nozzle is one of the locations that will be reevaluated for environmental assisted fatigue, and the reanalysis will consider the effects of potential rapid cycling as necessary. This action will be completed under the Fatigue Monitoring Program. As such, the effects of fatigue on the feedwater nozzles will be managed for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(iii).

Reactor Vessel Internals A general assembly drawing of the important RPV internals components is shown in UFSAR Figure 3.9-8. Fatigue evaluations for the reactor vessel internals were performed as part of the internals design. The fatigue analyses of the reactor vessel internals are considered time-limited aging analyses because they are based on numbers of design cycles that were expected to occur in 40 years of operation. The fatigue analyses of the reactor vessel internals were reviewed during the extended power uprate.

GGNS will monitor transient cycles using the Fatigue Monitoring Program and assure that action is taken if any of the actual cycles approach their analyzed numbers. As such, the Fatigue Monitoring Program will manage the effects of aging due to fatigue on the reactor vessel internals in accordance with 10 CFR 54.21(c)(1)(iii).

Reactor Recirculation Pumps UFSAR Section 3.9.1.2.1.4 describes the Byron-Jackson recirculation pump fatigue analysis.

The fatigue analysis for the reactor recirculation pump casing considered the RCS fatigue transients specified by GE. The analysis justified exempting portions of the case from analysis and determined that the remaining locations met 1974 ASME Section III code fatigue requirements. Usage factors were calculated for several locations in the pump cover that were

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-29 later reanalyzed due to modifications to install shaft sleeves and modify the seal water heat exchanger.

The Fatigue Monitoring Program will manage the effects of aging due to fatigue on the reactor recirculation pumps in accordance with 10 CFR 54.21(c)(1)(iii).

Control Rod Drives The Class 1 portions of the control rod drives were analyzed for fatigue. The cumulative usage factors are low and the tracking of cycles under the Fatigue Monitoring Program ensures the fatigue on these components remains acceptable. The Fatigue Monitoring Program will manage the effects of aging due to fatigue on the control rod drives in accordance with 10 CFR 54.21(c)(1)(iii)

Class 1 Piping The piping specifications for GGNS identified that the ASME Class 1 piping must be analyzed for the transient cycle drawings provided by General Electric. Detailed fatigue analyses were then generated for GGNS to analyze multiple locations on each system within the ASME Class 1 boundary. The fatigue analyses of the Class 1 piping are considered time-limited aging analyses because they are based on numbers of design cycles that were expected to occur in 40 years of operation. GGNS will monitor the cycles actually incurred compared to the cycles analyzed using the Fatigue Monitoring Program and assure that action is taken if the actual cycles approach their analyzed numbers. As such, the Fatigue Monitoring Program will manage the effects of aging due to fatigue on the ASME Section III piping in accordance with 10 CFR 54.21(c)(1)(iii).

A.2.2.2 Non-Class 1 Metal Fatigue The design of ASME III Code Class 2 and 3 piping systems incorporates the Code stress reduction factor for determining acceptability of piping design with respect to thermal stresses. In general, 7000 thermal cycles are assumed, allowing a stress reduction factor of 1.0 in the stress analyses. GGNS evaluated the validity of this assumption for 60 years of plant operation. The results of this evaluation indicate that the 7000 thermal cycle assumption will not be exceeded for 60 years of operation. Therefore, the pipe stress calculations remain valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i).

Non-class 1 components other than piping system components require fatigue analyses if they were built to a section of the code such as ASME Section III, NC-3200 or ASME Section VIII, Division 2. A review of the non-Class 1 components other than piping identified non-Class 1 fatigue analysis applicable to expansion joints. Design specifications and calculations were identified for expansion joints with fatigue analyses for a bounding number of cycles, which were identified as time limited aging analyses. Evaluation of these analyses determined the number of

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-30 cycles were adequate for 60 years of operation. Therefore, the non-Class 1 expansion joint TLAAs are valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i).

A.2.2.3 Effects of Reactor Water Environment on Fatigue Life NUREG/CR-6260 addresses the application of environmental factors to fatigue analyses (CUFs) and identifies locations of interest for consideration of environmental effects. Section 5.6 of NUREG/CR-6260 identified the following component locations to be the most sensitive to environmental effects for newer vintage General Electric plants. These locations and the subsequent calculations are directly relevant to GGNS.

(1) Reactor vessel shell and lower head (2) Reactor vessel feedwater nozzle (3) Reactor recirculation piping (including inlet and outlet nozzles)

(4) Core spray line reactor vessel nozzles and associated Class 1 piping (5) Residual heat removal nozzles and associated Class 1 piping (6) Feedwater line Class 1 piping To support the license renewal application GGNS performed a screening evaluation of these six locations using the guidance provided in NUREG-1801 revision 2. This screening has determined there are locations that when the current usage factor is increased to account for the environmental effects, the resulting usage is greater than 1. Prior to the period of extended operation GGNS will update the fatigue usage calculations using refined fatigue analyses to determine valid CUFs less than 1.0 when accounting for the effects of reactor water environment.

This includes applying the appropriate Fen factors to valid CUFs determined using an NRC-approved version of the ASME code or NRC-approved alternative (e.g., NRC-approved code case). GGNS will review design basis ASME Class 1 component fatigue evaluations to determine whether the locations that have been evaluated for the effects of the reactor coolant environment on fatigue include the limiting component within the reactor coolant pressure boundary. Environmental effects on fatigue for these critical components may be evaluated using one of the following sets of formulae:

Carbon and Low Alloy Steels Those provided in NUREG/CR-6583, using the applicable ASME Section III fatigue design curve.

Those provided in Appendix A of NUREG/CR-6909, using either the applicable ASME Section III fatigue design curve or the fatigue design curve for carbon and

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-31 low alloy steel provided in NUREG/CR-6909 (Figures A.1 and A.2, respectively, and Table A.1).

A staff-approved alternative.

Austenitic Stainless Steels Those provided in NUREG/CR-5704, using the applicable ASME Section III fatigue design curve.

Those provided in NUREG/CR-6909, using the fatigue design curve for austenitic stainless steel provided in NUREG/CR-6909 (Figure A.3 and Table A.2).

A staff-approved alternative.

Nickel Alloys Those provided in NUREG/CR-6909, using the fatigue design curve for austenitic stainless steel provided in NUREG/CR-6909 (Figure A.3 and Table A.2).

A staff-approved alternative.

If an acceptable CUF cannot be calculated, GGNS will repair or replace the affected locations before exceeding an environmentally adjusted CUF of 1.0.

GGNS will manage the effects of fatigue, including environmentally assisted fatigue, under the Fatigue Monitoring Program for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(iii).

A.2.3 Environmental Qualification of Electrical Components The GGNS Environmental Qualification (EQ) of Electric Components Program implements the requirements of 10 CFR 50.49 (as further defined by the Division of Operating Reactors Guidelines, NUREG-0588, and Reg. Guide 1.89). The program requires action before individual components exceed their qualified life. In accordance with 10 CFR 54.21(c)(1)(iii),

implementation of the EQ Program provides reasonable assurance that the effects of aging on components associated with EQ time-limited aging analyses will be adequately managed such that the intended functions can be maintained for the period of extended operation.

A.2.4 Fatigue of Primary Containment, Attached Piping, and Components Grand Gulf utilizes a BWR Mark III containment. As described in UFSAR Section 3.8.1.3, the containment was initially designed in accordance with the loads defined in GE Topical Report NEDO 11314-08 (GESSAR Appendix 3B). Additional loads initially defined in GE document 22A4365, Interim Containment Loads Report (ICLR), Rev. 2 and later defined by GE document 22A7000, Rev. 2 (GESSAR II, Appendix 3B; Grand Gulf FSAR, Appendix 6D), have been considered for the final design verification of the containment.

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-32 UFSAR Appendix 6A, Section 3BA.7.2.2.3 (page 6A-14) identifies the quenchers were designed for a conservatively high value of 18,000 cycles of fatigue.

UFSAR Section 3.8.1.4.2 identifies the analysis of the suppression pool and cylinder wall liner plate. Fatigue analysis for the suppression pool and cylinder wall liner plate was performed using subsections NE and NB of the ASME Code,Section III, Division I, 1971 Edition with Summer of 1973 Addenda.

As shown on UFSAR Figure 3.6A-33, the guard pipe assemblies utilize bellows. Calculations were identified for the bellows on the guard pipe assemblies that analyzed a large number of cycles of flexure due to normal operation and earthquakes and are therefore considered TLAAs.

As shown on UFSAR Figure 9.1-15 and described in UFSAR Section 9.1.4.2.3.11, the GGNS fuel transfer tube also uses bellows. A calculation was identified for the bellows on the transfer tube that analyzed a large number of cycles of flexure due to normal operation or earthquakes and is therefore considered a TLAA.

GGNS will monitor transient cycles using the Fatigue Monitoring Program and assure that action is taken if any of the actual cycles approach their analyzed numbers. As such, the Fatigue Monitoring Program will manage the effects of aging due to fatigue on the primary containment in accordance with 10 CFR 54.21(c)(1)(iii).

A.2.5 Other Plant-Specific TLAA A.2.5.1 Erosion of the Main Steam Line Flow Restrictors GGNS UFSAR Section 5.4.4.4 identifies for the stainless steel main steam flow restrictors, "Only very slow erosion will occur with time." The section later postulates that even with an erosion rate of 0.004 inches per year, the increase in choked flow after 40 years would be no more than five percent. This was evaluated as a TLAA.

Entergy Corporation evaluated the erosion-corrosion rate for the main steam flow elements. This analysis considered the specific material present in GGNS flow restrictors and determined the expected erosion-corrosion rate when operating at the velocities that would be present following EPU. The evaluation determined the expected erosion-corrosion rate would be much less than the conservative value in the UFSAR. Using this value, the expected total erosion after 60 years would remain less than the conservative total erosion value identified in the UFSAR for 40 years.

This analysis has been projected through the period of extended operation in accordance with 10CFR54.21(c)(1)(ii).

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-33 A.2.5.2 Determination of Intermediate High-Energy Line Break Locations UFSAR Section 3.6A.2 identifies for GGNS that the determinations of intermediate high-energy line break locations included an evaluation based on CUFs being less than 0.1 if other stress criteria are also met. The usage factors, as calculated in the design fatigue analyses, account for the design transients assumed for the original 40-year life of the plant. Therefore, the determination of cumulative usage factors used in the selection of postulated high-energy line break locations is considered a TLAA.

The Fatigue Monitoring Program will identify when the transients for high-energy piping systems are approaching their analyzed numbers of cycles. If the design cycles indicate the cycle limit for exceeding a CUF of 0.1 will be exceeded, the design calculations for that system will be reviewed to determine if any additional locations should be designated as postulated high-energy line breaks. If other locations are determined to require consideration as postulated break locations, actions will be taken to address the new break locations.

The determination of intermediate high-energy line break locations is considered a TLAA that is dispositioned by 10 CFR 54.21(c)(1)(iii). The program that will manage this is the Fatigue Monitoring Program.

A.2.5.3 Fluence Effects for the Reactor Vessel Internals The design specification 22A4052 for the reactor vessel internals components includes requirements beyond the ASME design requirements for austenitic stainless steel base metal components exposed to greater than 1 x 1021 nvt (> 1 MEV) or weld metal greater than 5 x 1020 nvt (> 1 MEV).

Entergy Corporation performed a fluence analysis of the components included in the design specification 22A4052 at EPU operating conditions for 60 years plant life (54 EFPY). Location-specific fluence levels were determined. The internal core support structure components were then evaluated against the irradiation criteria in the design specification. The results of the evaluation were that the GGNS internal core support structure components meet the design specification at EPU operating conditions for 54 EFPY.

Therefore, this analysis has been projected through the period of extended operation in accordance with 10CFR54.21(c)(1)(ii).

Grand Gulf Nuclear Station License Renewal Application Technical Information Appendix A Updated Safety Analysis Report Supplement Page A-34 A.3 REFERENCES A.3-1

[GGNS License Renewal Applicationlater]

A.3-2

[NRC SER for GGNS License Renewallater]

A.3-3 Letter GNRO-2010/00056, from Michael Krupa, Entergy, to USNRC, License Amendment Request, Extended Power Uprate, Grand Gulf Nuclear Station, Unit 1, Dated September 8, 2010.

A.3-4

[NRC SER for GGNS EPUlater]