ML20309A744

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2 to Updated Final Safety Analysis Report, Chapter 18, Aging Management Programs and Activities
ML20309A744
Person / Time
Site: Mcguire, McGuire  Duke Energy icon.png
Issue date: 10/08/2020
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Duke Energy Carolinas
To:
Office of Nuclear Reactor Regulation
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RA-19-0424
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McGuire Nuclear Station UFSAR Chapter 18 Table of Contents 18.0 Aging Management Programs and Activities 18.1 Introduction 18.1.1 References 18.2 Aging Management Programs and Activities 18.2.1 Alloy 600 Program 18.2.1.1 Deleted Per 2020 Update.

18.2.1.2 Control Rod Drive Mechanism Nozzle and Other Vessel Closure Penetration Inspection 18.2.1.3 Pressurizer Inspection 18.2.2 Borated Water Systems Stainless Steel Inspection 18.2.3 Bottom-Mounted Instrumentation Thimble Tube Inspection Program 18.2.4 Chemistry Control Program 18.2.5 Containment Inservice Inspection Plan - IWE 18.2.6 Deleted Per 2005 Update 18.2.7 Crane Inspection Program 18.2.8 Fire Protection Program 18.2.8.1 Sprinkler Branch Lines 18.2.8.2 Main Fire Pump Strainer 18.2.8.3 Jockey Pump Strainer 18.2.8.4 Tank and Connected Piping 18.2.8.5 Turbine Building Manual Hose Stations 18.2.8.6 Turbine Building Strainer Inspections 18.2.8.7 Fire Protection System Selective Leaching Inspections 18.2.9 Flood Barrier Inspection 18.2.10 Flow Accelerated Corrosion Program 18.2.11 Boric Acid Corrosion Control Program 18.2.12 Galvanic Susceptibility Inspection 18.2.13 Heat Exchanger Activities 18.2.13.1 Component Cooling Heat Exchangers 18.2.13.2 Containment Spray Heat Exchangers 18.2.13.3 Diesel Generator Engine Cooling Water Heat Exchangers 18.2.13.4 Control Area Chilled Water 18.2.13.5 Pump Motor Air Handling Units 18.2.13.6 Pump Oil Coolers 18.2.14 Ice Condenser Engineering Inspection 18.2.15 Inaccessible Non-EQ Medium-Voltage Cables Program 18.2.16 Inservice Inspection Program 18.2.16.1 McGuire Unit 1 Cold Leg Elbow 18.2.16.2 Small Bore Piping 18.2.16.3 Control Rod Drive Mechanism Penetration Thermal Sleeve Inspection 18.2.17 Inspection Program For Civil Engineering Structures and Components 18.2.18 Liquid Waste System Inspection 18.2.19 Non-EQ Insulated Cables and Connections Program 18.2.20 Pressurizer Spray Head Examination 18.2.21 Preventive Maintenance Activities 18.2.21.1 Condenser Circulating Water System Internal Coating Inspection 18.2.21.2 Refueling Water Storage Tank Internal Coating Inspection 18.2.21.3 Nuclear Service Water System Strainer Elements Inspection 18.2.21.4 Auxiliary Feedwater Storage Tank Internal Coating Inspection 18.2.21.5 Reactor Makeup Water Storage Tank Internal Coating Inspection (13 APR 2020) 18 - i

McGuire Nuclear Station UFSAR Chapter 18 18.2.22 Reactor Vessel Integrity Program 18.2.23 Reactor Vessel Internals Program 18.2.24 Selective Leaching Inspection 18.2.25 Service Water Piping Corrosion Program 18.2.26 Sump Pump Systems Inspection 18.2.27 Treated Water Systems Stainless Steel Inspection 18.2.28 Underwater Inspection of Nuclear Service Water Structures 18.2.29 Ventilation Area Pressure Boundary Sealants Inspection 18.2.30 Waste Gas System Inspection 18.2.31 Battery Rack Inspection 18.2.32 Steam Generator Surveillance Program 18.3.33 Additional Chemistry Commitment - Visual Inspection of Auxiliary Feedwater and Main Feedwater Piping 18.3.34 Fuse Holder Inspection 18.3.35 Boral Monitoring Program 18.3.36 Non-EQ High Range Radiation and Neutron Flux Instrumentation Circuits Program 18.3.37 Buried Piping Integrity (NEI 09-14) Program 18.2.38 References 18.3 Additional Commitments 18.3.1 Battery Rack Inspections (Moved to section 18.2.31 per 2020 update) 18.3.2 Steam Generator Surveillance Program (Moved to section 18.2.32 per 2020 update) 18.3.3 Additional Chemistry Commitment - Visual Inspection of Auxiliary Feedwater and Main Feedwater Piping (Moved to section 18.2.33 per 2020 update) 18.3.4 Fuse Holder Program (Moved to section 18.2.34 per 2020 update) 18.3.5 Boral Monitoring Program (Moved to section 18.2.35 per 2020 update) 18.4 Newly Identified SSCs 18.4.1 MNS Reviews for Newly Identified SSCs 18.4.1.1 Nuclear Service Water System Strainer Elements 18.4.1.2 Liquid Waste System Piping - Control Room Air Handling Unit Drains 18.4.1.3 Boral Spent Fuel Rack Neutron Attenuation Material 18.4.1.4 Earthen Dike on the North Perimeter of the McGuire Nuclear Station Site 18.4.1.5 Auxiliary Feedwater Storage Tanks 18.4.1.6 Main Condenser Hotwell and Condensate System Piping 18.4.1.7 Upper Surge Tanks and Condensate Storage System Piping 18.4.1.8 Reactor Makeup Water Storage Tanks and Reactor Makeup Water Pumps 18.4.1.9 Interior Fire Protection System Turbine Building Strainers 18.4.1.10 Radwaste Facility 18.4.1.11 Equipment Staging Building 18.4.1.12 Equipment Staging Building Ventilation System 18.4.1.13 Turbine Buildings and Service Building Fire Rated Assemblies 18.4.1.14 Turbine Buildings and Service Building Fire Rated Assemblies 18.4.1.15 Control Rod Drive Mechanism Penetration Thermal Sleeves 18.5 Newly Identified TLAAs 18.5.1 Upflow Modification TLAAs (13 APR 2020) 18 - ii

McGuire Nuclear Station UFSAR Chapter 18 List of Tables Table 18-1. Summary Listing of the Programs, Activities and TLAA Table 18-2. Deleted Per 2014 Update (13 APR 2020) 18 - iii

McGuire Nuclear Station UFSAR Chapter 18 THIS PAGE LEFT BLANK INTENTIONALLY.

(13 APR 2020) 18 - iv

McGuire Nuclear Station UFSAR Chapter 18 18.0 Aging Management Programs and Activities THIS IS THE LAST PAGE OF THE TEXT SECTION 18.0.

(13 APR 2020) 18.0 - 1

UFSAR Chapter 18 McGuire Nuclear Station THIS PAGE LEFT BLANK INTENTIONALLY.

18.0 - 2 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 18.1 Introduction Duke Energy Corporation prepared an Application for Renewed Operating Licenses of McGuire Nuclear Station, Units 1 and 2 and Catawba Nuclear Station, Units 1 and 2 (Application)

[Reference 1]. The application, including information provided in additional correspondence, provides sufficient information for the NRC to complete their technical and environmental reviews and provides the basis for the NRC to make the findings required by §54.29 (Final Safety Evaluation Report - Final SER) [Reference 2]. Pursuant to the requirements of

§54.21(d), the UFSAR supplement for the facility must contain a summary description of the programs and activities for managing the effects of aging and the evaluation of time-limited aging analyses for the period of extended operation determined by §54.21 (a) and (c),

respectively.

As an aid to the reader, Table 18-1 provides a summary listing of the programs, activities and time-limited aging analyses (TLAA) (topics) required for license renewal. The first column of Table 18-1 provides a listing of these topics. The second column of Table 18-1 indicates where the topic is located in the Application. This is an historical reference. The third column of Table 18-1 identifies where the description of the Program, Activity, or TLAA is located in either the McGuire UFSAR or in the McGuire Improved Technical Specifications (ITS).

Section 18.2 contains summary descriptions of the aging management programs and periodic inspections that are ongoing through the duration of the operating licenses of McGuire Nuclear Station.

Station documents will be established, implemented, and maintained to cover the aging management programs and activities described in Chapter 18.

The Corrective Action Program (AD-PI-ALL-0100) provides a structured approach for a formal corrective action program which facilitates the prioritization, evaluation, and correction of conditions adverse to quality, as defined by 10 CFR Part 50, Appendix B. This same corrective action program is credited for systems, structures, and components whose aging will be managed by the aging management programs and activities described herein.

McGuire Nuclear Station has completed reviews meeting the requirements of 10 CFR 54.37(b),

and determined that certain systems, structures, and components are in the scope of license renewal and newly identified.

Deleted per 2020 update .

Section 18.4 presents the evaluation of these newly identified systems, structures and components, as well as the aging management review development and the aging management activities that have been prescribed to ensure that intended functions will be maintained through the period of extended operation. Section 18.5 presents the evaluation of newly identified TLAAs associated with the 40-60 year license renewal period of extended operation.

18.1.1 References

1. M. S. Tuckman (Duke) letter dated June 13, 2001, to Document Control Desk (NRC), Application to Renew the Operating Licenses of McGuire Nuclear Station, Units 1 and 2, and Catawba Nuclear Station, Units 1 and 2, Docket Nos. 50-369, 50-370, 50-413, and 50-414.

(13 APR 2020) 18.1 - 1

UFSAR Chapter 18 McGuire Nuclear Station

2. NUREG-1772, Safety Evaluation Report Related to the License Renewal of McGuire Nuclear Station, Units 1 and 2, and Catawba Nuclear Station, Units 1 and 2, Docket Nos.

50-369, 50-370, 50-413, and 50-414, March 2003.

THIS IS THE LAST PAGE OF THE TEXT SECTION 18.1.

18.1 - 2 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 18.2 Aging Management Programs and Activities 18.2.1 Alloy 600 Program The purpose of the Alloy 600 Aging Management Review, as proposed during the license renewal review process for McGuire Nuclear Station, was to ensure that nickel-based alloy locations are adequately inspected by the Inservice Inspection Plan or other existing programs, to demonstrate the general oversight and management of cracking due to primary water stress corrosion cracking (PWSCC) in susceptible materials. Implementation of the Alloy 600 Aging Management Review was completed, as required, prior to June 12, 2021 (the end of the initial license of McGuire Unit 1).

The comprehensive Alloy 600 Aging Management Program supersedes the Alloy 600 Aging Management Review and meets the intent of MRP-126, Generic Guidance for Alloy 600 Management (EPRI), as a mandatory initiative for NEI 03-08, Guideline for the Management of Materials Issues. The Alloy 600 Aging Management Program provides the basis for inspections of susceptible materials in accordance with ASME Section XI and applicable code cases with conditions set forth in 10CFR50.55a. The program is updated as necessary to reflect any new or revised commitments made by Duke Energy in response to industry operating experience or NRC communications and requirements related to Alloy 600.

Deleted paragraph(s) per 2020 update.

Deleted paragraph(s) per 2018 update.

Deleted paragraph(s) per 2020 udpate.

Deleted Per 2012 Update.

18.2.1.1 Deleted per 2020 update Deleted paragraph(s) per 2020 update.

18.2.1.2 Control Rod Drive Mechanism Nozzle and Other Vessel Closure Penetration Inspection Deleted paragraph(s) per 2020 update Deleted paragraph(s) per 2018 update.

The scope of the Control Rod Drive Mechanism and Other Vessel Closure Penetration Inspection includes the control rod drive mechanism nozzles, head vent penetrations and carbon steel head surface of each reactor vessel as described in the ASME Section XI Code, including applicable Code Cases, subject to the conditions imposed by 10CFR Part 50.55a.

These penetrations include 78 Control Rod Drive Mechanism (CRDM) type penetrations, and one head vent penetration. The Control Rod Drive Mechanism and Other Vessel Closure Penetration Inspection also includes the four auxiliary head adaptors (AHAs) located on the outer portion of the reactor vessel head at 0°, 90°, 180°, and 270°, in accordance with the ASME Section XI Code, including applicable Code Cases, subject to the conditions imposed by 10 CFR Part 50.55a.

During inspection, if indications or evidence of leakage is detected, actions will be taken to disposition the condition as required by ASME Section XI and applicable code cases. Flaws which can be justified for continued service will be addressed through the Corrective Action (13 APR 2020) 18.2 - 1

UFSAR Chapter 18 McGuire Nuclear Station Program and in accordance with the ASME Section XI Code, including applicable Code Cases subject to the conditions imposed by 10CFR Part 50.55a.

Implementation of the Control Rod Drive Mechanism and Other Vessel Closure Penetration Inspection was completed prior to June 12, 2021 (the end of the initial license of McGuire Unit 1). The Control Rod Drive Mechanism and Other Vessel Closure Penetration Inspection is implemented under the controls of the Alloy 600 Aging Management Program.

18.2.1.3 Pressurizer Inspection Deleted paragraph(s) per 2020 udpate.

The Pressurizer inspection will detect cracking of pressurizer connections containing Alloy 600/82/182 materials prior to loss of component intended function. The program performs inspections, as applicable, including volumetric examination of any structural weld overlays per ASME Section XI and ASME Code Case N-770 requirements, as conditioned by 10 CFR 50.55a.

Evidence of leakage will be addressed in accordance with the Boric Acid Corrosion Control Program, including evaluation by engineering to determine extent of condition and applicability to other locations. The station Corrective Action Program will be utilized to evaluate the need for additional NDE methods and increased inspection scopes, including like locations and other Duke units. If circumferential cracking is observed in either the pressure boundary or non-pressure boundary portions of any locations covered under the scope of NRC Bulletin 2004-01, Duke will develop plans to perform an adequate extent-of-condition evaluation and Duke will discuss those plans with cognizant NRC technical staff prior to restarting the affected unit.

The Pressurizer Inspection is implemented under the controls of the Alloy 600 Aging Management Program.

18.2.2 Borated Water Systems Stainless Steel Inspection Deleted paragraph(s) per 2020 update.

The Borated Water Systems Stainless steel inspection is a one-time inspection to manage cracking and loss of material in stainless components exposed to alternate wetting and drying borated water environment in the Containment Spray and Refueling Water Systems.

The Borated Water Systems Stainless Steel Inspection was completed prior to June 12, 2021, using a volumetric technique to inspect stainless steel piping, welds, and heat affected zones in the Containment Spray System in the area of the internal air/water interface at one of twelve locations of concern. The results of the inspection found no unacceptable loss of material or cracking that could result in a loss of the component intended function(s). The results of this inspection are considered to be bounding and are applied to stainless steel components exposed to an alternate wetting and drying borated water environment in the Refueling Water System.

18.2.3 Bottom-Mounted Instrumentation Thimble Tube Inspection Program Deleted paragraph(s) per 2020 update.

The Bottom Mounted Instrumentation Thimble Tube Inspection Program monitors tube wall degradation (wear) of the BMI thimble tubes. Inspection of the BMI thimble tubes is performed using eddy current testing. All accessible thimble tubes are inspected. The frequency of examination is based on an analysis of the data obtained using wear rate relationships that are 18.2 - 2 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 predicted based on Westinghouse research that is presented in WCAP-12866, Bottom Mounted Instrumentation Flux Thimble Wear [Reference 2]. The eddy current results are trended and inspections are planned prior to the refueling outage in which thimble tube wear is predicted to exceeding the Acceptance Criteria, below. This ensures that the thimble tubes continue to perform their pressure boundary function.

The acceptance criterion for the BMI thimble tubes is 80% through wall, as developed in Westinghouse WCAP-12866, Bottom Mounted Instrumentation Flux Thimble Wear, and reported to the NRC by Duke [Reference 1.] Thimble tubes that are predicted to exceed the acceptance criterion may be capped or repositioned. Specific corrective actions and confirmatory actions are implemented in accordance with the corrective action program.

18.2.4 Chemistry Control Program The purpose of the Chemistry Control Program is to manage loss of material and/or cracking of components exposed to borated water, closed cooling water, fuel oil, and treated water environments. This program manages the relevant conditions that lead to the onset and propagation of loss of material, cracking, and fouling which could lead to a loss of structure or component intended functions. Relevant conditions are specific parameters such as halogens, dissolved oxygen, conductivity, biological activity, and corrosion inhibitor concentrations that could lead to loss of material and/or cracking if not properly controlled.

The Chemistry Control Program contains system specific acceptance criteria that are based on the guidance provided in EPRI PWR Primary Water Chemistry Guidelines, EPRI PWR Secondary Water Chemistry Guidelines, and EPRI Closed Cooling Water Chemistry Guideline.

18.2.5 Containment Inservice Inspection Plan - IWE The Containment Inservice Inspection Plan - IWE was developed to implement applicable requirements of 10 CFR 50.55a. Section 50.55a(g)(4) requires that throughout the service life of nuclear power plants, components which are classified as either Class MC or Class CC pressure retaining components and their integral attachments must meet the requirements, except design and access provisions and preservice examination requirements, set forth in Section XI of the ASME Code and Addenda that are incorporated by reference in §50.55a(b).

Furthermore, §50.55a(g)(4)(v)(A) requires that metal containment pressure retaining components and their integral attachments must meet the inservice inspection, repair, and replacement requirements applicable to components which are classified as ASME Code Class MC. These requirements are subject to the limitation listed in paragraph (b)(2)(vi) and the modifications listed in paragraphs (b)(2)(viii) and (b)(2)(ix) of §50.55a, to the extent practical within the limitations of design, geometry and materials of construction of the components

[Reference 3].

18.2.6 Deleted Per 2005 Update 18.2.7 Crane Inspection Program Deleted paragraph(s) per 2020 update.

The Crane Inspection Program detects aging effects through visual examination of the crane rails and girders, including seismically restrained cranes. Examination and assessment of the condition of a structure is performed using guidance provided in codes and standards such as:

(13 APR 2020) 18.2 - 3

UFSAR Chapter 18 McGuire Nuclear Station

  • ANSI B30.2.0, Overhead and Gantry Cranes, American National Standard, Section 2-2, Safety Standards for Cableways, Cranes, Derricks, Hoists, Hooks, Jacks and Slings, The American Society of Mechanical Engineers, New York.
  • ANSI B30.16, Overhead Hoists (Underhung), The American Society of Mechanical Engineers, New York.
  • 29 CFR Chapter XVII, 1910.179, Occupational Safety and Health Administration, Overhead and Gantry Cranes The Crane Inspection Program is implemented by plant procedures and through the work management system using model work orders. The acceptance criterion is no unacceptable visual indication of loss of material. Structures and components that do not meet the acceptance criteria are evaluated by engineering for continued service and repaired as required. Structures and components which are deemed unacceptable are documented under the Corrective Action Program. Specific corrective actions and confirmatory actions are implemented in accordance with the Corrective Action Program.

18.2.8 Fire Protection Program Elements of the Fire Protection Program that serve to manage aging are implemented in accordance with Selected Licensee Commitments as identified in Table 18-1. Enhancements to the Fire Protection Program to provide surveillances for sprinkler branch lines, main fire pump strainer, jockey pump strainer, tank and connected piping and turbine buildi9ng manual hose stations were completed prior to June 12, 2021 (the end of the initial license of McGuire Unit 1).

18.2.8.1 Sprinkler Branch Lines The integrity of the sprinkler branch lines is assured by sprinkler flow tests performed by procedure every 18 months. Additionally, fouling of sprinkler branch lines that do not receive flow during periodic testing is managed by a combination of wall thickness measurements under the Service Water Piping Corrosion, and periodic surveys of fire protection distribution piping to investigate for evidence of fouling.

Additionally, a sample of sprinklers will be either inspected or replaced at 50 years of operation in accordance with NFPA 25.

18.2.8.2 Main Fire Pump Strainer The Main Fire Pump Strainer Inspection is a periodic inspection to identify any loss of material of each main fire pump strainer. Raw water flow could result in loss of material to the strainer.

The acceptance criteria for the Main Fire Pump Strainer Inspection is no unacceptable loss of material that could result in a loss of component intended function(s) as determined by engineering evaluation.

18.2.8.3 Jockey Pump Strainer The Jockey Pump Strainer Inspection is a periodic inspection to identify any loss of material of each jockey fire pump strainer basket. Raw water flow could result in loss of material to the strainer. The acceptance criteria for the Jockey Pump Strainer Inspection is no unacceptable loss of material that could result in a loss of component intended function(s) as determined by engineering evaluation.

18.2 - 4 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 18.2.8.4 Tank and Connected Piping The Tank and Connected Piping Internal Inspection is a periodic inspection to manage loss of material of the internal surfaces of the carbon steel fire protection system pressure maintenance accumulator tank and connecting piping and valves supplying high-pressure air. The internal carbon steel surfaces of the tank are coated with an epoxy coating. Continued presence of an intact coating precludes loss of material of the internal surfaces of the carbon steel tank. This preventive maintenance activity inspects the internal coating of the fire protection system pressure maintenance accumulator tank to check the condition of the coating to identify coating failures and the condition of the connecting piping supplying highpressure air to identify loss of material. High-pressure air system piping connected to the tank is also monitored under the Service Water Piping Corrosion Program. The Tank and Connected Piping Internal Inspection is a condition monitoring activity.

18.2.8.5 Turbine Building Manual Hose Stations All of the hose stations in the Turbine Building within the scope of license renewal are periodically tested every three (3) years by opening each hose station valve to verify no flow blockage. Additionally, periodic surveys of fire protection distribution piping are performed to investigate for evidence of fouling.

18.2.8.6 Turbine Building Strainer Inspections The Turbine Building Fire Protection Strainer Inspections is a periodic inspection of strainers in Interior Fire Protection System Strainers located in the Turbine Building at least once every 10 years to identify degradation due to loss of material and/or fouling. The acceptance criteria for the Turbine Building Fire Protection Strainer Inspections is no unacceptable loss of material and/or fouling that could result in a loss of component intended function(s) as determined by engineering evaluation.

18.2.8.7 Fire Protection System Selective Leaching Inspections The Fire Protection System Selective Leaching Inspections manages selective leaching of gray cast iron components in the Interior and Exterior Fire Protection Systems. The program was developed as a result of one-time inspections performed during the initial license period, which identified that selective leaching of gray cast iron components was occurring and required aging management through the PEO. The program inspects a sampling of gray cast iron components (valves, pump casings, strainer housings) to monitor degradation due to selective leaching. The frequency for this inspection is based on operating experience, to be performed at least once every 10 years. The program also includes periodic inspection of the main fire pumps on a 10 year frequency, during PMs for rebuild of those pumps.

Inspections for selective leaching will be performed by individuals experienced in detection of this aging mechanism, and will typically utilize the services of the McGuire Materials Laboratory.

All destructive examinations shall be done by individuals experienced in performance of metallurgical examinations. Acceptance criteria will be engineering evaluation that intended functions will be maintained through the next inspection interval. The engineering evaluation will apply inspection results to the overall population to determine if sample size or inspection frequency needs to be adjusted. Results that do not meet the acceptance criteria are addressed in the Corrective Action Program.

(13 APR 2020) 18.2 - 5

UFSAR Chapter 18 McGuire Nuclear Station 18.2.9 Flood Barrier Inspection The Flood Barrier Inspection manages cracking and change in material properties of the elastomeric flood seals to ensure that safety-related equipment is protected from floods and flooding flow paths such that no equipment safety-related intended functions or station safe shutdown capability are adversely impacted. This activity includes periodic visual inspections of the flood seals to identify degradation that could result in loss of the intended function of the flood seals. The Flood Barrier Inspection is a condition monitoring program.

18.2.10 Flow Accelerated Corrosion Program Deleted paragraph(s) per 2020 update.

The Flow Accelerated Corrosion Program is credited for managing loss of material due to flow accelerated corrosion of carbon steel piping, valves and cavitating venturies within the scope of license renewal and determined to be susceptible to flow accelerated corrosion. The program is consistent with the basic guidelines and recommendations provided by EPRI document NSAC-202L [Reference 5]. Component wall thickness is measured using volumetric examinations such as ultrasonic testing and radiography. Visual examinations are also employed when access to interior surfaces is allowed by component design. Component wall thickness acceptability is judged in accordance with the McGuire component design code of record. Inspection results are monitored and trended to determine the calculated rate of material loss, to detect impact from changes in operating or chemistry conditions, and schedule for the next inspection. If the calculated component wall thickness at the time of the next outage is projected to be less than the allowable minimum wall thickness with safety margin under the component design code of record, then the component will be repaired or replaced prior to system start-up. The as-inspected component can also be justified for continued service through additional detailed engineering analysis. Specific corrective actions are implemented in accordance with the Flow Accelerated Corrosion Program and the Corrective Action Program.

18.2.11 Boric Acid Corrosion Control Program Deleted paragraph(s) per 2020 update.

The purpose of the Boric Acid Corrosion Control Program is to detect boric acid intrusion and/or loss of material due to boric acid wastage prior to loss of structure or component intended function. The scope of the Boric Acid Corrosion Control Program includes electrical, mechanical, and structural components within the scope of license renewal that are located in the Auxiliary and Reactor Buildings where exposure to leaks from borated water systems is possible. Mechanical and structural components constructed of carbon steel, low alloy steel, and other susceptible materials are also included within the scope of the program.

The program consists of (a) visual inspection of external surfaces that are potentially exposed to borated water leakage, (b) timely discovery of leak path and removal of the boric acid residues, (c) assessment of the damage, and (d) follow-up inspection for adequacy. The Boric Acid Corrosion Control Program was implemented in response to NRC GL 88-05 and recent operating experience. Specific corrective actions are implemented in accordance with Boric Acid Corrosion Control Program procedures and the Corrective Action Program.

18.2.12 Galvanic Susceptibility Inspection Deleted paragraph(s) per 2020 udpate.

18.2 - 6 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 The Galvanic Susceptibility Inspection is a one-time inspection that manages loss of material due to galvanic corrosion of piping components McGuire Systems exposed to gas, unmonitored treated water, and raw water environments in the following systems:

  • Containment Ventilation Cooling Water System
  • Diesel Generator Room Sump Pump System
  • Exterior Fire Protection System
  • Interior Fire Protection System
  • Waste Gas System
  • Diesel Generator Engine Cooling Water System heat exchangers (channel heads, tubesheets)
  • Reciprocating Charging Pump Fluid Drive Oil Coolers (channel covers)

The galvanic couples within systems in the inspection are carbon steel, cast iron, and ductile iron (anodes) coupled to copper alloys or stainless steel (cathodes) and copper alloys (anodes) coupled to stainless steel (cathode).

The Galvanic Susceptibility Inspection was implemented in accordance with controlled plant procedures, and inspected a select set of carbon steel-stainless steel couples at McGuire using a volumetric examination technique. In some instances, visual examination was used when access to internal surfaces became available. Since inspection of all couples is impractical, inspections were performed at carbon steel -stainless steel couples, where galvanic corrosion is more likely to occur. A sentinel population of carbon steel-stainless steel couples located in raw water systems was inspected to determine whether loss of material due to galvanic corrosion will be an aging effect of concern for the period of extended operation. The results of this inspection were applied to other galvanic couples in the systems in the scope of the program.

The Galvanic Susceptibility Inspection was completed prior to June 12, 2021 (the end of the initial license of McGuire Unit 1).

The Galvanic Susceptibility Inspection determined that galvanic corrosion was occurring in bimetallic welded connections, such that ongoing aging management was required. The Service Water Piping Corrosion Program has been expanded to include monitoring for loss of material in this welded connections through the period of extended operation. The Service Water Piping Corrosion Program is also credited with ongoing inspections of the Diesel Generator Engine Cooling Water System heat exchangers for corrosion, including galvanic corrosion.

MNS is abandoning the Reciprocating Charging Pumps and auxiliary equipment (including the fluid drive oil coolers), with this modification to be completed prior to entering the period of extended operation for the respective units (June 12, 2021 for Unit 1, and March 3, 2023 for Unit 2). Subsequently, these SSCs will no longer perform an intended function, and therefore will not be subject to aging management during the period of extended operation.

Operating experience identified galvanic corrosion was occurring at brazed joints (carbon steel to copper) in the cooling water supply to pump motor air handling units in the Auxiliary Building (13 APR 2020) 18.2 - 7

UFSAR Chapter 18 McGuire Nuclear Station Ventilation System. These brazed connections are being eliminated by a design change, which will be completed prior to entering the period of extended operation for the respective unit (June 12, 2021 for Unit 1, and March 3, 2023 for Unit 2). Therefore, these air handling units will not require aging management for galvanic corrosion during the period of extended operation.

18.2.13 Heat Exchanger Activities Enhancements to the Heat Exchanger Preventive Maintenance Activities to provide surveillances for pump motor air handling units and pump oil coolers were completed prior to June 12, 2021 (the end of the initial license of McGuire Unit 1).

18.2.13.1 Component Cooling Heat Exchangers The purpose of the Performance Testing Activities - Component Cooling Heat Exchangers is to manage fouling of admiralty brass heat exchanger tubes that are exposed to raw water. The Performance Testing Activities - Component Cooling Heat Exchangers is a performance monitoring program that monitors specific component parameters to detect the presence of fouling which can affect the heat transfer function of the component.

The purpose of the Heat Exchanger Preventive Maintenance Activities - Component Cooling is to manage loss of material for parts of the component cooling heat exchanger exposed to raw water. The Heat Exchanger Preventive Maintenance Activities- Component Cooling is a condition monitoring program that monitors specific component parameters to detect the presence and assess the extent of material loss that can affect the pressure boundary function.

The program is credited with managing loss of material for admiralty brass, carbon steel, and stainless steel materials. Criteria such as ASME Code requirements, additional inspection results, and operating experience may be used to assess the severity of the degradation and the need for corrective actions.

18.2.13.2 Containment Spray Heat Exchangers The purpose of the Performance Testing Activities - Containment Spray Heat Exchangers is to manage fouling of stainless steel and titanium heat exchanger tubes that are exposed to raw water. The Performance Testing Activities - Containment Spray Heat Exchangers is a performance monitoring program that monitors specific component parameters to detect the presence of fouling, which can affect the heat transfer function of the component. Heat exchanger 2NSHX0004 has been replaced with a design having raw water on the tube side, which can be readily inspected and cleaned. Accordingly, heat transfer of this heat exchanger is maintained by a regular program of cleaning and inspection, consistent with the requirements of GL 89-13.

The purpose of the Heat Exchanger Preventive Maintenance Activities - Containment Spray is to manage loss of material for parts of the containment spray heat exchanger exposed to raw water. The Heat Exchanger Preventive Maintenance Activities - Containment Spray is a condition monitoring program that monitors specific component parameters to detect the presence and assess the extent of material loss that can affect the pressure boundary function.

The program is credited with managing loss of material for stainless steel and titanium materials. Criteria such as ASME Code requirements, additional inspection results, and operating experience may be used to assess the severity of the degradation and the need for corrective actions.

18.2 - 8 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 18.2.13.3 Diesel Generator Engine Cooling Water Heat Exchangers The purpose of the Performance Testing Activities - Diesel Generator Engine Cooling Water Heat Exchangers is to manage fouling of admiralty brass heat exchanger tubes that are exposed to raw water. The Performance Testing Activities - Diesel Generator Engine Cooling Water Heat Exchangers is a performance monitoring program that monitors specific component parameters to detect the presence of fouling, which can affect the heat transfer function of the component.

The purpose of the Heat Exchanger Preventive Maintenance Activities Diesel Generator Engine Cooling Water is to manage loss of material for parts of the diesel generator engine cooling water heat exchanger exposed to raw water. The Heat Exchanger Preventive Maintenance Activities Diesel Generator Engine Cooling Water is a condition monitoring program that monitors specific component parameters to detect the presence and assess the extent of material loss that can affect the pressure boundary function. The program is credited with managing the subject aging effects for admiralty brass heat exchanger tubes. Criteria such as ASME Code requirements, additional inspection results, and operating experience may be used to assess the severity of the degradation and the need for corrective actions.

18.2.13.4 Control Area Chilled Water The purpose of the Heat Exchanger Preventive Maintenance Activities - Control Area Chilled Water is to manage fouling and loss of material of parts of the control room area chillers exposed to raw water. The Heat Exchanger Preventive Maintenance Activities - Control Area Chilled Water is a condition monitoring program that monitors specific component parameters to detect the presence and assess the extent of material loss that can affect the pressure boundary functions and periodically cleans the chiller tubes to manage fouling. The Heat Exchanger Preventive Maintenance Activities - Control Area Chilled Water is credited for managing loss of material or fouling for carbon steel and copper-nickel alloy materials. Criteria such as ASME Code requirements, additional inspection results, and operating experience may be used to assess the severity of the degradation and the need for corrective actions.

18.2.13.5 Pump Motor Air Handling Units The purpose of Heat Exchanger Preventive Maintenance Activities - Pump Motor Air Handling Units is to manage loss of material and fouling of copper heat exchanger tubes that are exposed to raw water. The Heat Exchanger Preventive Maintenance Activities - Pump Motor Air Handling Units is a new condition monitoring program that will detect the presence and assess the extent of material loss that can affect the pressure boundary function and will periodically dP test/clean the heat exchanger tubes to manage fouling. The scope of Heat Exchanger Preventive Maintenance Activities - Pump Motor Air Handling Units is the tubes in the following McGuire heat exchangers of the Auxiliary Building Ventilation System:

  • Fuel Pool Cooling Pump Motor Air Handling Units Criteria such as ASME Code requirements, additional inspection results, and operating experience may be used to assess the severity of the degradation and the need for corrective actions. A destructive examination was performed on one of the twelve cooling units within the scope of the program prior to June 12, 2021 (the end of the initial license of McGuire Unit 1).

(13 APR 2020) 18.2 - 9

UFSAR Chapter 18 McGuire Nuclear Station This examination identified no significant loss of material and fouling of the tubes of the pump motor air handling units in the Auxiliary Building Ventilation System.

18.2.13.6 Pump Oil Coolers The purpose of Heat Exchanger Preventive Maintenance Activities - Pump Oil Coolers is to manage loss of material and fouling of copper-nickel heat exchanger tubes that are exposed to raw water. The Heat Exchanger Preventive Maintenance Activities - Pump Oil Coolers is a new condition monitoring program that monitors specific component parameters to detect the presence and assess the extent of material loss that can affect the pressure boundary function and periodically cleans the heat exchanger tubes to manage fouling. While fouling is managed currently by periodic cleaning, this comprehensive program to manage both loss of material and fouling is a new plant program for license renewal. The scope of Heat Exchanger Preventive Maintenance Activities - Pump Oil Coolers is the tubes in the following McGuire heat exchangers of the Nuclear Service Water System:

  • Centrifugal Charging Pump Bearing Oil Cooler
  • Centrifugal Charging Pump Speed Reducer Oil Cooler
  • Reciprocating Charging Pump Bearing Oil Cooler
  • Reciprocating Charging Pump Fluid Drive Oil Cooler
  • Safety Injection Pump Bearing Oil Cooler Criteria such as ASME Code requirements, additional inspection results, and operating experience may be used to assess the severity of the degradation and the need for corrective actions. A non-destructive examinationn (Eddy Current Test) was performed on 100% of the tubes of one of the sixteen coolers within the scope of the program prior to June 12, 2021 (the end of the initial license of McGuire Unit 1). The results of this inspection were used to establish ongoing eddy current test requirements for all the coolers in the scope of the program.

The Heat Exchanger Preventive Maintenance Activities - Pump Oil Coolers includes periodic inspection of the Pump Fluid Drive Oil Cooler tube sheet for selective leaching.

The Reciprocating Charging Pumps will be abandoned prior to the period of extended operation, such that the Reciprocating Charging Pump Bear Oil Cooler and Fluid Drive Oil Cooler no longer perform an intended function, are not in the scope of license renewal and will not require ongoing aging management during the period of extended operation.

18.2.14 Ice Condenser Engineering Inspection The Ice Condenser Engineering Inspection manages loss of material due to corrosion of the steel structural components in the ice condenser environment. The Ice Condenser Engineering Inspection includes periodic visual inspections of the ice condenser upper plenum, lower plenum, and top deck blankets to identify degradation that could impact the ability of the ice condenser to perform its intended function. The Ice Condenser Engineering Inspection is a condition monitoring program.

18.2.15 Inaccessible Non-EQ Medium-Voltage Cables Aging Management Program Deleted paragraph(s) per 2020 update.

18.2 - 10 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 The Inaccessible Non-EQ Medium-Voltage Cables Aging Management Program periodically tests medium-voltage cables within the scope of the program to provide an indication of the condition of the conductor insulation. The scope of the Inaccessible Non-EQ Medium-Voltage Cables Aging Management Program includes inaccessible non-EQ medium-voltage cables within the scope of 10 CFR 54.4 that are exposed to significant voltage and to standing water (for any period of time).

Medium-voltage cables within the scope of the Inaccessible Non-EQ Medium-Voltage Cables Aging Management Program are tested at least once every 10 years to provide an indication of the condition of the conductor insulation. The specific type of test performed will be determined before each test and will be a proven test for providing an indication of the condition of the conductor insulation related to aging effects caused by moisture and voltage stress. Further investigation is performed through the Corrective Action Program when the acceptance criteria are not met. Initial cable testing under the Inaccessible Non-EQ Medium-Voltage Cables Aging Management Program was completed by June 12, 2021 (the end of the initial license of McGuire Unit 1).

The Inaccessible Non-EQ Medium-Voltage Cables Aging Management Program includes annual inspections of sumps and sump pumps to prevent accumulation of water and cable submergence in manholes.

18.2.16 Inservice Inspection Plan The McGuire Inservice Inspection Plan implements the requirements of 10 CFR 50.55a for Class 1, 2, and 3 components and Class 1, 2, 3, and MC component supports. The examinations are performed to the extent practicable within the limitations of design, geometry and materials of construction of the component. The period of extended operation for McGuire will contain the 5th and 6th ten-year inservice inspection intervals.

The Inservice Inspection Plan includes the following inspections and activities:

  • McGuire Unit 1 Cold Leg Elbow
  • Examination of the reactor vessel internals clevis insert fasteners. (Note - Duke has submitted and the NRC has approved a revision to the McGuire Reactor Vessel Internals Inspection Program based on the requirements of MRP-227-A [References 28 and 29].)
  • Inspection of Pressurizer Surge Line welds and the Safety Injection Nozzle welds credited in flaw tolerance evaluations for aging management of cracking due to environmentally assisted fatigue (EAF).

18.2.16.1 McGuire Unit 1 Cold Leg Elbow Reduction in fracture toughness due to thermal embrittlement can be an aging effect for certain types of cast austenitic stainless steel in locations where temperatures continuously exceed 482ºF. In a May 19, 2000 letter to NEI, Christopher I. Grimes, Chief License Renewal and (13 APR 2020) 18.2 - 11

UFSAR Chapter 18 McGuire Nuclear Station Standardization Branch clarified that not all cast austenitic stainless steels are subject to thermal embrittlement [Reference 8]. The piping components and reactor coolant pumps fabricated from cast austenitic stainless steel were evaluated using the acceptance criteria set forth in the above letter. For those components requiring evaluation, only the McGuire 1, 27 1/2-inch ID Loop B cold leg elbow exceeds the NRC-established threshold and is susceptible to thermal embrittlement which requires aging management for license renewal.

The McGuire Unit 1 27 1/2-inch ID Loop B cold leg elbow is fabricated from SA-351 CF8, was statically cast, and contains no niobium. The elbow is the only piping item that exceeds the delta ferrite screening criterion, therefore, reduction of fracture toughness by thermal embrittlement is an aging effect requiring aging management for this elbow. The ferrite number is calculated at 22% using Hulls equivalent factors.

An augmented inspection with elements from Code Case N-481 will be used to manage reduction of fracture toughness by thermal embrittlement for the affected elbow during the period of extended operation. A VT-1 visual examination will be performed of the external surfaces of the welded joints that connect the affected elbow to adjacent piping segments prior to entering the period of extended operation. VT-1 inspections of the welded joints will be repeated in the fifth and sixth inspection intervals.

The initial VT-1 inspection has been completed and a detailed evaluation to demonstrate the safety and serviceability of the elbow has been performed. This evaluation concluded that, for the maximum predicted fatigue crack growth for McGuire Unit 1 Loop 2 cold leg elbow, the postulated flaw will not grow beyond the tolerable flaw size after 60-years of crack growth

[Reference 27].

Enhancements to the Inservice Inspection Program to provide surveillances for the Unit 1 Cold Leg Elbow Inspection were completed by June 12, 2021 (the end of the initial license of McGuire Unit 1).

Deleted paragraph(s) per 2020 update.

18.2.16.2 Small Bore Piping Small bore piping is defined as piping less than 4-inch NPS. Cracking has been identified as an aging effect requiring programmatic management for Reactor Coolant System small bore piping for the period of extended operation.

A set of susceptible small bore piping locations will be volumetrically examined on each unit.

The Program will sample 10% of the ASME Class 1 Small Bore piping welds that are high safety significance and are considered to have degradation mechanisms, for each unit, in both the 5th and 6th interval. Locations to be examined will be determined based on consideration of damage mechanisms. Damage mechanisms to be considered include fatigue, stress corrosion, and flow assisted corrosion/flow wastage. Cracking due to thermal fatigue resulting from stratification of fluids and turbulent penetration flow is an aging effect that will be addressed.

For McGuire, Small Bore Piping Examinations will be performed during each inservice inspection interval during the period of extended operation following issuance of renewed operating licenses for McGuire Nuclear Station.

18.2.16.3 Control Rod Drive Mechanism Penetration Thermal Sleeve Inspections Wear of the Control Rod Drive Mechanism Penetration Thermal Sleeves is an aging effect that can lead to interference with control rod movement. The Inservice Inspection Program includes 18.2 - 12 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 augmented inspections of the thermal sleeves to determine wear measurements, and trending to ensure that corrective actions are taken prior to failure of the thermal sleeve flange.

18.2.17 Inspection Program For Civil Engineering Structures and Components The Inspection Program for Civil Engineering Structures and Components is intended to meet the requirements of 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants (the Maintenance Rule). This program:

(1) monitors and assesses mechanical components, civil structures and components and their condition in order to provide reasonable assurance that they are capable of performing their intended functions in accordance with the current licensing basis. The program includes condition monitoring of ventilation area pressure boundary sealants; (2) includes nuclear safety-related structures which enclose, support, or protect nuclear safety-related systems and components, non-safety related structures whose failure may prevent a nuclear safety-related system or component from fulfilling its intended function, and non safety-related structures which support equipment relied on during certain regulated events.

NEI 96-03, Industry Guideline for Monitoring the Condition of Structures at Nuclear Power Plants, has been used as guidance in the preparation of the Inspection Program for Civil Engineering Structures and Components. Examination and assessment of the condition of a structure is performed using guidance provided in codes and standards such as:

  • ACI 349.3, Evaluation of Existing Nuclear Safety-Related Concrete Structures The program credits periodic inspections required by FERC and performed by the Hydro Group for aging management of the Lake Norman Dike located north of the plant.

Deleted paragraph(s) per 2020 update.

18.2.18 Liquid Waste System Inspection Deleted paragraph(s) per 2020 update.

The Liquid Waste System Inspection is a one time inspection credited for aging management of cast iron, stainless steel and carbon steel components exposed to unmonitored treated and borated water environments or raw water environments in the following McGuire systems:

  • Component Cooling System - At McGuire, the waste evaporator package consists of four heat exchangers, three of which are within the scope of license renewal. One of the four heat exchangers will be inspected. The inspection results will be applied to the other three stainless steel heat exchanger components exposed to unmonitored treated water environments.
  • Liquid Waste Recycle System - stainless steel components exposed to an unmonitored borated water environment
  • Liquid Waste System piping draining the Control Room Air Handling Units, up to the point the piping exits the Control Building (13 APR 2020) 18.2 - 13

UFSAR Chapter 18 McGuire Nuclear Station The Liquid Waste System inspection also includes a one time inspection of both Annulus loop seal drains, to address NRC concerns raised during review of the License Renewal Application regarding the environment at these locations.

The Liquid Waste System Inspection is intended to detect the presence and extent of:

  • loss of material due to crevice and pitting corrosion and cracking due to stress corrosion/intergranular attack in stainless steel components exposed to unmonitored borated and treated water environments.
  • loss of material due to crevice, pitting, microbiologically influenced corrosion and cracking due to stress corrosion in stainless steel components exposed to raw water environments.
  • loss of material due to crevice, general, pitting, and microbiologically influenced corrosion in carbon steel and cast iron components exposed to raw water environments.

The Liquid Waste System one time inspection was completed prior to June 12, 2021 (the end of the initial license of McGuire Unit 1).The Liquid Waste System Inspection used volumetric examination techniques to inspect the material/environment combinations identified above. For the waste evaporator package heat exchangers in the Component Cooling System, destructive metallurgical investigation was used due to access limitations associated with heat exchanger configuration.

The Liquid Waste System one time inspection determined that most SSCs crediting this program showed little evidence of age related degradation, such that ongoing aging management is not required. The Annulus loop seal drains were the sole exception, with one of the loop seals determined to have incurred wall thinning such that ongoing aging management will be prescribed. Both annulus loop seal drains have been referred to the Service Water Piping Corrosion Program for ongoing aging management through the period of extended operation.

18.2.19 Non-EQ Insulated Cables and Connections Aging Management Program Deleted paragraph(s) per 2020 update.

The Non-EQ Insulated Cables and Connections Aging Management Program is an ongoing program to detect aging effects for accessible non-EQ insulated cables and connections caused by heat and radiation prior to loss of intended function. The Non-EQ Insulated Cables and Connections Aging Management Program includes accessible (able to be approached and viewed easily) non-EQ (not subject to 10 CFR 50.49 Environmental Qualification requirements) insulated electrical cables and connections (power, instrumentation and control applications) installed in the Reactor Buildings, Auxiliary Building and Turbine Building.

Accessible non-EQ insulated cables and connections installed in the Reactor Buildings, Auxiliary Building and Turbine Building are visually inspected at least once every 10 years per the Non-EQ Insulated Cables and Connections Aging Management Program for cable and connection jacket surface anomalies such as embrittlement, discoloration, cracking or surface contamination. Cable and connection jacket surface anomalies are precursor indications of conductor insulation aging degradation from heat or radiation in the presence of oxygen andmay indicate the existence of an adverse localized equipment environment. An adverse localized equipment environment is a condition in a limited plant area that is significantly more severe than the specified service condition for the insulated cable or connection.

18.2 - 14 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 The acceptance criterion for the Non-EQ Insulated Cables and Connections Aging Management Program is no unacceptable visual indications of cable and connection jacket surface anomalies that suggest conductor insulation degradation exists, as determined by engineering evaluation.

An unacceptable indication is defined as a noted condition or situation that, if left unmanaged, could lead to a loss of the intended function. When an adverse localized equipment environment is identified for a cable or connection, a determination is made as to whether the same condition or situation is applicable to other accessible or inaccessible cables or connections. Further investigation through the corrective action program is performed when the acceptance criteria are not met.

Initial inspections under the Non-EQ Insulated Cables and Connections Aging Management Program were completed by June 12, 2021 (the end of the initial license of McGuire Unit 1).

18.2.20 Pressurizer Spray Head Examination The McGuire Nuclear Station transition to a fire protection program based on NFPA 805 removed the licensing requirement to attain cold shutdown within a prescribed time frame in the event of a fire, instead requiring the plants to attain and maintain "safe and stable conditions",

which can be achieved at hot standby. Accordingly, the pressurizer spray heads are no longer functionally required for compliance with fire protection regulations. Since the pressurizer spray heads are not credited in other regulated events in 10 CFR 54.4(a)(3), or for any design basis events, and have no potential adverse interactions with safety related SSCs or safety functions, it follows that they no longer perform an intended function, and therefore are no longer within the scope of license renewal. Since the pressurizer spray heads are no longer in the scope of license renewal, there are no ongoing aging management requirements under 10 CFR 54, and the Pressurizer Spray Head Examination activity has been deleted.

18.2.21 Preventive Maintenance Activities 18.2.21.1 Condenser Circulating Water System Internal Coating Inspection The Preventive Maintenance Activities - Condenser Circulating Water System Internal Coating Inspection manages loss of material and cracking that could lead to loss of pressure boundary function. The internal carbon steel surfaces of the large diameter intake and discharge piping in the Condenser Circulating Water System are coated to prevent the raw water environment from contacting the internal surfaces. Continued presence of an intact coating precludes loss of material of the internal surfaces of the carbon steel intake and discharge piping. This inspection will periodically check the condition of the coating and look for coating degradation. The Preventative Maintenance Activities - Condenser Circulating Water System Internal Coating Inspection also periodically inspects the Condenser Circulating Water System crossover piping for loss of material and fouling.

The Preventive Maintenance Activities - Condenser Circulating Water System Internal Coating Inspection was also formerly credited with managing loss of material and cracking of the external surfaces of components in the underground environment by providing symptomatic evidence of the condition of the other buried piping external surfaces. This approach to condition monitoring is not considered to provide sufficient assurance of effectiveness in light of site and industry operating experience, and has been replaced with implementation of the MNS Buried Piping Integrity (NEI 09-14) Program for aging management of external surfaces of buried piping and tanks in the scope of license renewal.

(13 APR 2020) 18.2 - 15

UFSAR Chapter 18 McGuire Nuclear Station 18.2.21.2 Refueling Water Storage Tank Internal Coating Inspection The purpose of the Preventive Maintenance Activities - Refueling Water Storage Tank Internal Coating Inspection is to manage loss of material of the internal surfaces of the carbon steel refueling water storage tanks. The internal carbon steel surfaces of the refueling water storage tank are coated with a phenolic epoxy paint that prevents borated water and air from contacting the internal surfaces. Continued presence of an intact coating precludes loss of material of the internal surfaces of the carbon steel refueling water storage tank that could lead to loss of pressure boundary function. This preventive maintenance activity inspects the internal coating of the refueling water storage tanks to check the condition of the coating and to identify coating failures. The Preventive Maintenance Activities - Refueling Water Storage Tank Internal Coating Inspection is a condition monitoring program.

18.2.21.3 Nuclear Service Water System Strainer Elements Inspection The purpose of the Preventive Maintenance Activities - Nuclear Service Water System Strainer Elements Inspection is to manage loss of material / fouling of the stainless steel Nuclear Service Water System strainer elements. This preventive maintenance activity periodically inspects the strainer elements to verify physical integrity is not degraded by age related degradation and fouling is not occurring. The Preventive Maintenance Activities - Nuclear Service Water System Strainer Elements Inspection is a condition monitoring program.

18.2.21.4 Auxiliary Feedwater Storage Tank Internal Coating Inspection The purpose of the Preventive Maintenance Activities - Auxiliary Feedwater Storage Tank Internal Coating Inspection is to manage loss of material of the internal surfaces of the carbon steel Auxiliary Feedwater Storage Tanks. The internal carbon steel surfaces of the Auxiliary Feedwater Storage Tanks are coated with an epoxy paint that prevents air from contacting the internal surfaces. Continued presence of an intact coating precludes loss of material of the internal surfaces of the carbon steel Auxiliary Feedwater Storage Tanks that could lead to loss of pressure boundary function. This preventive maintenance activity inspects the internal coating of the Auxiliary Feedwater Storage Tanks to check the condition of the coating and to identify coating failures. The Preventive Maintenance Activities - Auxiliary Feedwater Storage Tank Internal Coating Inspection is a condition monitoring program.

18.2.21.5 Reactor Makeup Water Storage Tank Internal Coating Inspection The purpose of the Preventive Maintenance Activities - Reactor Makeup Water Storage Tank Internal Coating Inspection is to manage loss of material of the internal surfaces of the carbon steel Reactor Makeup Water Storage Tanks. The internal carbon steel surfaces of the Reactor Makeup Water Storage Tanks are coated with an epoxy paint that prevents air from contactingthe internal surfaces. Continued presence of an intact coating precludes loss of material of the internal surfaces of the carbon steel Reactor Makeup Water Storage Tanks that could lead to loss of pressure boundary function. This preventive maintenance activity inspects the internal coating of the Reactor Makeup Water Storage Tanks to check the condition of the coating and to identify coating failures. The Preventive Maintenance Activities - Reactor Makeup Water Storage Tank Internal Coating Inspection is a condition monitoring program.

18.2.22 Reactor Vessel Integrity Program Deleted paragraph(s) per 2020 update.

The Reactor Vessel Integrity Program monitors reduction of fracture toughness of reactor vessel beltline materials by irradiation embrittlement to determine the need for operating restrictions on 18.2 - 16 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 the inlet temperature, neutron spectrum, and neutron flux. The Reactor Vessel Integrity Program includes all reactor vessel beltline materials as defined by 10 CFR 50.61(a)(3). The cavity dosimetry provides a method for verification of fast neutron exposure distribution within the reactor vessel beltline region and establishes a mechanism to enable long term monitoring of neutron exposure. The program incorporates:

  • Submittal of reports required by 10 CFR Part 50 Appendix H which include a capsule withdrawal schedule, a summary report of capsule withdrawal and test results within one year of capsule withdrawal and if needed a date when a Technical Specification change will be made to change pressure-temperature limits or procedures to meet pressure temperature limits. As surveillance capsules are withdrawn and either tested or stored, documentation will be updated accordingly and submitted to the NRC in accordance with 10 CFR 50, Appendix G.
  • Pressure-Temperature curves are maintained in the plant Technical Specifications.

18.2.23 Reactor Vessel Internals Inspection Deleted paragraph(s) per 2020 udpate.

Reactor Vessel Internals Inspection commitments arising from license renewal were developed based on industry knowledge available at that time. These commitments included an allowance that permits Duke Energy to modify or eliminate these inspections based on industry data or other evaluations if plant specific justification is provided to demonstrate the basis for the modification or elimination.

The program relies on implementation of the inspection and evaluation guidelines in MRP-227-A and MRP-228 to manage the aging effects on the reactor vessel internal components. This program is used to manage: (a) cracking, including stress corrosion cracking, primary water stress corrosion cracking, irradiation-assisted stress corrosion cracking, and cracking due to fatigue/cyclical loading; (b) loss of material induced by wear; (c) loss of fracture toughness due to either thermal aging, neutron irradiation embrittlement, or void swelling; (d) dimensional changes due to void swelling or distortion; and (e) loss of preload due to thermal and irradiation enhanced stress relaxation or creep.

By letter dated December 13, 2017 [Reference 28], Duke Energy submitted the Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to implement MRP-227-A. This submittal includes the information identified in Section 3.5.1 of the NRC Safety Evaluation on MRP-227, Revision 0, and therefore meets the requirement for application of MRP-227-A as a strategy for managing age-related material degradation in reactor vessel internal components. By letter dated March 8, 2019 the NRC provided the safety evaluation and approved for the Reactor Vessel Internals Aging Management Program based on MRP-227-A.

Deleted paragraph(s) per 2020 update.

RVI inspections are performed as augmented inspections. If inspection results of the inspection are not acceptable, then actions will be taken to repair or replace the affected items or to determine by analysis the acceptability of the items. Specific corrective actions and confirmation are implemented in accordance with the corrective action program.

(13 APR 2020) 18.2 - 17

UFSAR Chapter 18 McGuire Nuclear Station The Control Rod Drive Mechanism thermal sleeve flanges were determined to be newly identified based on operating experience that found flange wear might cause interference with Control Rod movement. Inspections of the Control Rod Drive Mechanism thermal sleeve flanges performed as augmented inspections under the Inservice Inspection Program are credited with management of this aging mechanism.

18.2.24 Selective Leaching Inspection The Selective Leaching Inspection is a one-time inspection that will detect the presence and extent of any loss of material due to selective leaching of brass and cast iron components exposed to raw water in the following McGuire systems:

  • Diesel Generator Room Sump Pump
  • Exterior Fire Protection
  • Groundwater Drainage
  • Interior Fire Protection
  • Condenser Circulating Water The Selective Leaching Inspection monitors the hardness of the wetted surface of cast iron pump casings and brass valve bodies in a raw water environment. Selective leaching is the dissolution of one metal in an alloy at the metal surface which leaves a weakened network of corrosion products that can be revealed by a Brinnell Hardness check or equivalent as reduction in material hardness.

The Selective Leaching Inspection was implemented prior to June 12, 2021. The inspection utilized investigations performed by the McGuire Metallurgy Unit to detect the presence of selective leaching. For cast iron, the initial scope of the inspection was to be one cast iron pump casing in the Exterior Fire Protection System at McGuire. Based on operating experience, inspections were expanded to components in each of the systems having gray cast iron components in the scope of the program. The Selective Leaching Inspection also included inspections of a sample of brass valves at McGuire in the Interior Fire Protection System.

The results of the Selective Leaching Inspection determined that selective leaching was occurring in gray cast iron components in the Interior and Exterior Fire Protection Systems to the extent that ongoing aging management is warranted. Selective leaching was also detected in Condenser Circulating Water System valves, but was determined to not have the potential to impact intended functions over the period of extended operation based on the wall thickness of the valves involved. Operating experience review also identified the occurrence of selective leaching in admiralty brass heat exchanger tubes exposed to raw water.

As a result of these findings, ongoing aging management of selective leaching in will be implemented through the period of extended operation as follows:

  • Main Fire Pumps subject to periodic refurbishment and replacement under the Fire Protection Program - Mechanical Component Inspections.
  • Component Cooling Water HX and Diesel Generator Cooling Water HX admiralty brass tubes subject to eddy current testing under the Heat Exchangers Preventive Maintenance Activities 18.2 - 18 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 Selective leaching was also identified in the Reciprocating Charging Pump Fluid Drive Oil Cooler tube sheet. The Reciprocating Charging Pump and Auxiliaries (including the Fluid Drive Oil Coolers) will be abandoned prior to the period of extended operation. Subsequently, they will no longer perform an intended function, and therefore not be in the scope of license renewal.

18.2.25 Service Water Piping Corrosion Program Deleted paragraph(s) per 2020 update.

The MNS License Renewal Application credits the Service Water Piping Corrosion Program with managing loss of material such as that due to general corrosion, or particulate erosion that may occur in areas of higher flow velocity. The program also detects loss of material due to localized corrosion due to crevice, pitting, and microbiologically-influenced corrosion (MIC). The MNS License Renewal Application credits the Service Water Piping Corrosion Program with aging management for components in the following systems:

  • Containment Ventilation Cooling Water System
  • Exterior Fire Protection System
  • Interior Fire Protection System
  • Diesel Generator Cooling Water System
  • Control Area Chilled Water System
  • Safety Injection System
  • Chemical and Volume Control System The Service Water Piping Corrosion Program was expanded during license renewal implementation to manage loss of material in the following SSCs:
  • Diesel Generator Room Sump Pump System piping (from the Sump Pump Systems Inspection)
  • Annulus drain loop seals in the Liquid Waste System (from the Liquid Waste System Inspection)
  • Station Air System supply piping to the Fire Protection Pressurization Tank The Service Water Piping Corrosion Program was also expanded as a result of the Galvanic Susceptibility Inspection, to provide ongoing management for loss of material due to galvanic corrosion. The expanded scope includes for bimetallic welds in the following systems:
  • Containment Ventilation Cooling Water System
  • Diesel Generator Room Sump Pump System
  • Exterior Fire Protection System (13 APR 2020) 18.2 - 19

UFSAR Chapter 18 McGuire Nuclear Station

  • Interior Fire Protection System
  • Waste Gas System
  • Groundwater Drainage System The expanded scope also includes bimetallic joints in the Diesel Generator Engine Cooling Water System heat exchangers subcomponents (channel heads, tubesheets) Monitoring for degradation is accomplished using ultrasonic test techniques, which may be supplemented by visual inspections of the inside of the piping if access to the interior surfaces is provided.

Monitoring of localized corrosion is additionally supplemented by exterior piping inspections that may reveal pinhole leaks caused by localized corrosion.

Localized corrosion due to pitting and microbiologically influenced corrosion (MIC) may reveal itself through pinhole leaks in the piping components. The geometry of pinholes means that they are not a structural integrity concern. Further, in open cycle raw water systems these pinhole leaks typically do not lead to loss of the component intended function, since sufficient flow at prescribed pressures can still be provided by the system. These localized concerns will lead structural integrity concerns only when a significant number of pinholes are present. When indications of a pinhole are found, volumetric wall thickness measurements are taken in the area. A trend of indications of through-wall leaks due to pitting corrosion or MIC provides evidence when localized corrosion may become a structural integrity concern and will trigger corrective actions by the Service Water Piping Corrosion Program.

Bimetallic connections managed by the program for galvanic corrosion are identified and subject to periodic inspection at a frequency that assures that intended functions will be maintained.

Inspections for galvanic corrosion may be sample based, with technical justification provided for the sample size and inspection schedule to provide assurance of program effectiveness.

The Service Water Piping Corrosion Program is implemented using controlled plant procedures and work orders. Acceptance criteria are based on design requirements. Specific corrective actions and confirmation are implemented in accordance with the corrective action program.

The program requires regular reviews of inspection results and operating experience, to ensure that program effectiveness is continually maintained.

The License Renewal Application identified loss of material due to corrosion as an aging effect requiring management for carbon steel subcomponents in the Reciprocating Charging Pump Bearing Oil Coolers, to be managed by the Service Water Piping Corrosion Program. MNS is abandoning the Reciprocating Charging Pumps and auxiliary equipment (including the Bearing Oil Coolers), with this modification be completed prior to entering the period of extended operation for the respective unit (June 12, 2021 for Unit 1, and March 3, 2023 for Unit 2).

Subsequently, these SSCs will no longer perform an intended function, and therefore will not be subject to aging management during the period of extended operation.

18.2.26 Sump Pump Systems Inspection Deleted paragraph(s) per 2020 update.

The Sump Pump Systems Inspection is a one-time inspection that manages loss of material due to crevice, general, pitting and microbiologically influenced corrosion in a limited set of mechanical components constructed of carbon steel, cast iron, and stainless steel exposed to sump environments in the following McGuire systems:

  • Diesel Generator Room Sump Pump System 18.2 - 20 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18

  • Groundwater Drainage System.

As described in the license renewal application, the Diesel Generator Room Sump Pump System was selected for inspection because the system contains a representation of all of the materials present within the other sump environments. The sump environment in the Diesel Generator Room Sump Pump System is a potential combination of leakage of raw water, fuel oil, and treated water. Inspection of the Diesel Generator Room Sump Pump System provides a representative review of the condition of mechanical component materials subject to a sump environment. Inspection locations will be at piping low points, pump casings, and valve bodies where materials are continuously wetted by the raw water environment or subject to alternate wetting and drying. The results of this inspection will be applied to the mechanical components in the Groundwater Drainage System.

The Sump Pump Systems one-time inspection was completed prior to June 12,2021. The Sump Pump Systems Inspection inspected sump pump components located within the Diesel Generator Room Sump Pump System using volumetric and destructive examination techniques.

A sump pump casing was removed and subjected to metallurgical investigation and engineering evaluation, which concluded that the sump pump casings should be suitable for continued service through the period of extended operation. Inspection of sump pump system piping determined that carbon steel piping components in the sump pump environment are subject to corrosion, and require ongoing aging management. Accordingly, carbon steel piping in the Diesel Generator Room Sump Pump System and the Groundwater Drainage System has been referred to the Service Water Piping Corrosion Program for ongoing monitoring through the period of extended operation.

18.2.27 Treated Water Systems Stainless Steel Inspection Deleted paragraph(s) per 2020 update.

The Treated Water Systems Stainless Steel Inspection is a one-time inspection to detect the presence and extent of any loss of material or cracking of stainless steel components exposed to unmonitored treated water environments in the Nuclear Solid Waste Disposal System.

The Treated Water Systems Stainless Steel Inspections at McGuire inspected stainless steel components, welds, and heat affected zones, as applicable, in the McGuire Nuclear Solid Waste Disposal System. The McGuire Nuclear Solid Waste Disposal System components within the scope of license renewal is a mixture of unmonitored treated water and spent resins sluiced from demineralizers in various systems. The environment is expected to contain contaminants in excess of the limits below which a concern would not exist for cracking and loss of material in stainless steel. A concentration of any contaminants present would occur in areas of low flow or stagnant conditions. As a result, inspections were performed in stagnant and low flow lines around the spent resin storage tanks using volumetric techniques. In addition to the volumetric examination, a visual examination of the interior of a valve was conducted to determine the presence of pitting corrosion.

The Treated Water Systems Stainless Steel Inspections one-time inspection was completed prior to June 21, 2021. The Treated Water Systems Stainless Steel Inspections identified no unacceptable loss of material or cracking that could result in the loss of the component intended function(s) as determined by engineering evaluation. No ongoing aging management activities are needed to manage these aging effects in the Nuclear Solid Waste Disposal System.

(13 APR 2020) 18.2 - 21

UFSAR Chapter 18 McGuire Nuclear Station 18.2.28 Underwater Inspection of Nuclear Service Water Structures Deleted paragraph(s) per 2020 update.

The purpose of the Underwater Inspection of Nuclear Service Water Structures is to detect loss of material of steel components and loss of material and cracking of concrete components in the following structures:

  • Standby Nuclear Service Water Intake Structure The Underwater Inspection of Nuclear Service Water Structures will detect loss of material of steel components and loss of material and cracking of concrete components prior to loss of structure or component intended functions. The Underwater Inspection of Nuclear Service Water Structures detects aging effects through visual examination performed every five years at McGuire. Examination and assessment of the condition of a structure is performed using guidance provided in codes and standards such as:
  • ACI 349.3, Evaluation of Existing Nuclear Safety-Related Concrete Structures
  • ACI 201, Guide for Making a Condition Survey of Concrete in Service Prior inspection reports are reviewed to ensure implementation of recommended corrective actions. The acceptance criteria are no unacceptable visual indication of (1) loss of material for steel components and (2) loss of material and cracking for concrete components, as determined by the accountable engineer. The qualifications of the accountable engineer are in accordance with the guidance provided in NRC Regulatory Guide 1.127. Structures and components which do not meet the acceptance criteria are evaluated by the accountable engineer for continued service and repair, as required. Structures and components which are deemed unacceptable are documented under the corrective action program. Specific corrective actions and confirmatory actions, as needed, are implemented in accordance with the corrective action program.

18.2.29 Ventilation Area Pressure Boundary Sealants Inspection Deleted paragraph(s) per 2020 update.

The Ventilation Area Pressure Boundary Sealants Inspection is a one-time inspection to detect cracking or shrinkage of the ventilation area pressure boundary structural sealants. The scope of the Ventilation Area Pressure Boundary Sealants Inspection is the pressure boundary tructural sealants installed in the ventilation pressure boundary of the Control Room, ECCS Pump Room, Annulus, and Fuel Handling areas. Pressure boundary structural sealants include, but are not limited to, sealants in the interface between a structural wall, floor or ceiling and a non-structural component such as duct, piping, electrical cables, doors, and nonstructural walls.

The Ventilation Area Pressure Boundary Sealants Inspection performs visual inspections of a representative sample of structural sealants at each station. Locations of inspections are based on severity of the local ambient conditions taking into consideration temperature and radiation.

The sample locations selected provide a leading indication of the condition of all structural sealants within the scope of this activity. The acceptance criterion for the Ventilation Area Pressure Boundary Sealants Inspection is no unacceptable cracking or shrinking that could 18.2 - 22 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 result in the loss of the intended function of the structural sealant as determined by engineering evaluation.

The Ventilation Area Pressure Boundary Sealants Inspection one-time inspection was completed prior to June 12, 2021. Evaluation of inspection results determined that degradation of structural sealants was occurring, and should be subject to ongoing aging management.

Ongoing condition monitoring of structural sealants will be performed during periodic inspections under the Inspection Program for Civil Engineering Structures and Components.

18.2.30 Waste Gas System Inspection Deleted paragraph(s) per 2020 update.

The Waste Gas System Inspection is a one-time inspection to detect the presence and extent of any loss of material due to general, crevice, or pitting corrosion or cracking due to stress corrosion in carbon steel and stainless steel materials subject to an unmonitored treated water environment. The Waste Gas System Inspection will also detect the presence and extent of loss of material due to general corrosion in carbon steel materials subject to a gas environment.

The Waste Gas System Inspection uses a volumetric technique to inspect three sets of material/environment combinations. As an alternative, visual examinations are used should access to internal surfaces become available.

1) For carbon steel components exposed to unmonitored treated water environments at McGuire, inspections will be performed on the lower portions of decay tanks and associated drain lines where condensate is likely to accumulate. If no parameters are known that would distinguish the susceptible locations at McGuire, one of the eight available at McGuire will be examined based on accessibility and radiological concerns.

The results of this inspection will be applied to the remainder of the Waste Gas System carbon steel components within the scope of license renewal exposed to unmonitored treated water environment.

2) For stainless steel components exposed to unmonitored treated water environments at McGuire, inspections will be performed on the seal water path of the waste gas compressor. If no parameters are known that would distinguish the susceptible locations at McGuire, one of the two available at McGuire will be examined based on accessibility and radiological concerns. The results of this inspection will be applied to the remainder of the Waste Gas System stainless steel components within the scope of license renewal exposed to unmonitored treated water environment.
3) For the carbon steel components exposed to a gas environment at McGuire, an inspection will be performed on components within the scope of license renewal located between the volume control tanks and the waste gas compressor phase separators. If no parameters are known that would distinguish the most susceptible locations at McGuire, one location at McGuire will be examined based on accessibility and radiological concerns. The results of this inspection will be applied to the remainder of the Waste Gas System carbon steel components within the scope of license renewal exposed to gas environments.

The results of these inspections can be applied to the remainder of the carbon steel components exposed to gas environments. The acceptance criteria for the Waste Gas System Inspection is no unacceptable loss of material or cracking that could result in a loss of the component intended function(s) as determined by engineering evaluation. If engineering (13 APR 2020) 18.2 - 23

UFSAR Chapter 18 McGuire Nuclear Station evaluation determines that continuation of the applicable aging effects could cause a loss of component intended function(s) under current licensing basis design conditions for the period of extended operation, then specific corrective actions will be implemented in accordance with the Corrective Action Program.

The Waste Gas System Inspection was performed prior to June 12, 2021. Inspection results were determined to be acceptable, such that ongoing aging management is not required. Aging management of galvanic couples in the Waste Gas System was addressed under the Galvanic Susceptibility Inspection Commitment.

18.2.31 Battery Rack Inspections Battery rack inspections conducted in accordance with ITS SR 3.8.4.3, SLC 16.8.3.3, SLC 16.9.7.12, and SLC 16.9.7.16 shall include the structural supports and anchorages.

18.2.32 Steam Generator Surveillance Program The inspections of the Steam Generator Surveillance Program follow the requirements of Technical Specification 5.5.9 "Steam Generator (SG) Program" for ensuring that integrity is maintained for steam generator tubes and tube plugs. In addition, steam generator secondary side tube support components are monitored consistent with NEI 97-06.

18.2.33 Additional Chemistry Commitment The Additional Chemistry Commitment is a one time inspection of carbon steel components in the Auxiliary Feedwater System and Main Feedwater System components, intended to verify the effectiveness of chemistry controls. As described in regulatory correspondence, the program is based on opportunistic visual inspections of interior surfaces of the Auxiliary Feedwater System and Main Feedwater System.

The Additional Chemistry Commitment was completed, as required, prior to June 12, 2021. The review and evaluation of Additional Chemistry Control inspections found no instances wherein corrosion was identified that implicated chemistry controls are insufficient.

18.2.34 Fuse Holder Program For McGuire, Duke committed to implement the final version of the fuse holder interim staff guidance (initially provided to NEI by NRC letter dated May 16, 2002 and when finalized by the staff). This staff guidance was conveyed to the industry in LR-ISG-05, "Interim Staff Guidance on the Identification and Treatment of Electrical Fuse Holders for License Renewal", and subsequently incorporated into NUREG-1801, Generic Aging Lessons Learned (GALL) Report. Implementation of this commitment was satisfied by June 12, 2021 (the end of the initial license of McGuire Unit 1), by an evaluation of McGuire fuse holders that determined that an aging management program for fuse holders was not required.

18.2.35 Boral Monitoring Program The purpose of the Boral Monitoring Program is to assure that degradation of the Boral neutron absorbing material used in McGuire spent fuel pools that could compromise the criticality analysis will be detected. The Boral Monitoring Program relies on periodic inspection, testing, monitoring, and analysis to assure that the required 5% sub-criticality margin is maintained during the period of extended operation.

18.2 - 24 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 18.2.36 Non-EQ High Range Radiation and Neutron Flux Instrumentation Circuits Program The Non-EQ High-Range Radiation and Neutron Flux Instrumentation Circuits Program relies upon a proven cable test for detecting deterioration of the insulation system (such as insulation resistance tests, time domain reflectometry tests, or other testing judged to be effective in determining cable system insulation condition) for Non-EQ High-Range Radiation and Neutron Flux Instrumentation Circuit cables in the scope of the program. (Since the cables in the Non-EQ High-Range Radiation and Neutron Flux Instrumentation Circuits Program are disconnected during instrument calibration testing, a review of calibration results is not a viable method of condition monitoring.) Initial testing of cables in the scope of the program will be completed prior to the license renewal period of extended operation for the respective unit. Recurring inspections will be performed at a frequency based on engineering evaluation and will be performed at least once every 10 years.

18.2.37 Buried Piping Integrity (NEI 09-14)

The Buried Piping Integrity (NEI 09-14) Program is based on the requirements of NEI 09-14 to manage MNS buried piping assets, and specifically includes buried piping and tanks in the scope of license renewal. As a formal NSIAC initiative, the Buried Piping Integrity Guideline is recognized by the NRC as an industry commitment, and such is subject to inspection and enforcement as addressed in NRC Policy Issue "Industry Initiatives in the Regulatory Process",

presented in SECY-00-0116. The MNS License Renewal Application credited the Preventive Maintenance Activities - Condenser Circulating Water System Internal Coating Inspection with providing symptomatic information regarding condition monitoring of buried piping in the scope of license renewal. The Preventive Maintenance Activities - Condenser Circulating Water System Internal Coating Inspection Aging Management Program has been replaced by the Buried Piping Integrity Program to provide assurance that MNS buried piping and tanks in the scope of license renewal will perform its intended function through the period of extended operation.

18.2.38 References

1. M. S. Tuckman (Duke) letter dated July 30, 1991, NRC Bulletin 88-09, Thimble Tube Thinning in Westinghouse Reactors, McGuire Nuclear Station, Docket Nos. 50-369 and 50-370; Catawba Nuclear Station, Docket Nos. 50-413 and 50-414.
2. WCAP-12866, Bottom Mounted Instrumentation Flux Thimble Wear, January 1991.
3. 10 CFR Part 50, §50.55a, Codes and Standards.
4. W. T. Russell (NRC) letter dated November 19,1993 to William Rasin, (NUMARC),

Safety Evaluation for Potential Reactor Vessel Head Adapter Tube Cracking.

5. EPRI NSAC-202L-R2, Recommendations for an Effective Flow Accelerated Corrosion Program, Revision 2, April 1999.
6. Nuclear System Directive 104, Materiel/Condition/Housekeeping, Foreign Material Exclusion and Seismic Concerns, Revision 33.
7. Procedure AD-MN-ALL-0006, Fluid Leak Management, Revision 0.
8. C. I. Grimes (NRC) letter dated May 19, 2000 to D. J. Walters (NEI), License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components, Project No. 690.

(13 APR 2020) 18.2 - 25

UFSAR Chapter 18 McGuire Nuclear Station

9. Guideline for the Management of Adverse Localized Equipment Environments, EPRI, Palo Alto, CA: 1999. EPRI TR-109619.
10. McGuire Nuclear Station Updated Final Safety Analysis Report, as revised.
11. WCAP-14040, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, June 1994.
12. ASME Boiler and Pressure Vessel Code,Section III Nuclear Power Plant Components, Subsection ND Class 3 Components, 1971 edition.
13. NRC Bulletin 2003-02, Leakage from Reactor Vessel Lower Head Penetrations and Reactor Coolant Pressure Boundary Inegrity, August 21, 2003.
14. Deleted per 2018 update.
15. NRC Bulletin 2004-01, Inspection of Alloy 82/182/600 Materials Used in the Fabrication of Pressurizer Penetrations and Steam Space Piping Connections at PWRS, May 28, 2004.
16. Barron, Henry B. (Duke) to U. S. Nuclear Regulatory Commission, Duke Response to NRC Bulletin 2004-01, July 27, 2004.
17. McCollum, William R. (Duke) to U. S. Nuclear Regulatory Commission, Supplement to Response to NRC Bulletin 2004-01, September 21, 2004.
18. Barret, R. (NRC) to Marion, A. (NEI), Flaw Evaluation Guidelines, April 11, 2003.
19. PD-EG-PWR-1611, Boric Acid Corrosion Control Program (Program Description),

Rev 0.

20. Deleted per 2018 update.
21. Deleted per 2018 update.
22. Federal Register, 10 CFR Part 50, Industry Codes & Standards; Amended Requirements; Final Rule; September 10, 2008 - Pages 52742 and 52749.
23. Deleted per 2018 update.
24. Federal Register, 10 CFR Part 50, Incorporation by Reference of American Society of Mechanical Engineers Codes and Code Cases; Final Rule July 18, 2017.
25. AD-EG-PWR-1611, Boric Acid Corrosion Control Program-Implementation (Administrative Procedure), Rev. 0
26. Ray, Thomas D. (Duke) to U.S. Nuclear Regulatory Commission, Submittal of Information Related to Pressurizer Surge and Spray Nozzle Thermal Sleeves Attachments Welds for License Renewal, June 26, 2018
27. MCM 1201.01-1352.001, Flaw Tolerance Evaluation for Loop Two Cold Leg Elbow at McGuire Nuclear Station, Unit 1, Revision 0
28. Ray, Thomas D. (Duke) to U.S. Nuclear Regulatory Commission, Review Request for the Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227-A, December 13, 2017 18.2 - 26 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18

29. Ray, Thomas D. (Duke) to U.S. Nuclear Regulatory Commission, Resolution of Comments related to the Review Request for the Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227-A, May 9, 2018
30. EC 114606, Modifications to the Pressurizer Manways Due to Leakage Including (Optional) Use of Studs Instead of Bolts
31. NEI 97-06, Steam Generator Program Guidelines.

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McGuire Nuclear Station UFSAR Chapter 18 18.3 Additional Commitments 18.3.1 Battery Rack Inspections Moved to Section 18.2.31 per 2020 update.

18.3.2 Steam Generator Surveillance Program Moved to Section 18.2.32 per 2020 update.

18.3.3 Additional Chemistry Commitment - Visual Inspection of Auxiliary Feedwater and Main Feedwater Piping Moved to Section 18.2.33 per 2020 update.

18.3.4 Fuse Holder Program Moved to Section 18.2.34 per 2020 update.

18.3.5 Boral Monitoring Program Moved to Section 18.2.35 per 2020 update.

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McGuire Nuclear Station UFSAR Chapter 18 18.4 Newly Identified SSCs 18.4.1 McGuire Nuclear Station Reviews for Newly Identified SSCs Pursuant to the requirements of 10 CFR 54.37(b), McGuire Nuclear Station has performed reviews for newly identified SSCs, to be included in the scope of license renewal. These reviews have determined that newly identified SSCs do exist, subjected them to aging management reviews, and identified aging management programs to ensure that intended functions are maintained through the period of extended operation.

Consistent with the NRC's guidance in RIS 2007-16, the information that follows includes a description of newly identified SSCs at McGuire Nuclear Station, presents summary tables of aging management reviews, and then prescribes the aging management programs that will be used to manage the effects of aging. Where new McGuire aging management program activities are prescribed, a description is provided addressing the 10 elements of an effective aging management program, consistent with the information provided in Appendix B of the McGuire Nuclear Station License Renewal Application. Where an existing McGuire aging management program is credited for aging management of newly identified SSCs, but requires revision, the 10 element description of the program from the McGuire Nuclear Station License Renewal Application is provided, with revisions associated with newly identified SSCs incorporated in bold. If an existing McGuire aging management program is prescribed and no revision to the program is necessary to address newly identified SSCs, the program is identified in the aging management review summary, but the 10 element description of the program is not reprinted in this section.

18.4.1.1 Nuclear Service Water System Strainer Elements The Nuclear Service Water System Strainer Elements function to protect downstream safety-related SSCs from the effects of flow blockage due to fouling, and loss of heat transfer as a result of fouling. These strainer elements were considered non safety-related at the time the combined McGuire Nuclear Station, Units 1 and 2 and Catawba Nuclear Station, Units 1 and 2 License Renewal Application was approved, but have since been upgraded to safety-related to resolve site operating experience pertaining to biofouling. These strainer elements are now considered to perform an intended function, and are included in the scope of license renewal under 10 CFR 54.4(a)(1).

Results from a Mechanical aging management review identified that these strainer elements perform a filtration intended function, and could be susceptible to age related degradation due to loss of material and fouling. A new Preventive Maintenance Inspection Activity was prescribed to manage the effects of aging. A description of that program activity follows the aging management review summary table below.

An interdisciplinary review determined that there are no newly identified Civil / Structural or Electrical / I&C features associated with the Nuclear Service Water System Strainers.

A summary of the Mechanical aging management review for the Nuclear Service Water System strainer elements is provided in the following table:

(13 APR 2020) 18.4 - 1

UFSAR Chapter 18 McGuire Nuclear Station Component Component Material Internal Aging Aging Aging Type Function Environ. Effect Mechanism Management Program External Environ.

Crevice Corrosion Loss of Micro- Preventive Nuclear Material biologically Maintenance Service Influenced Activities Water Raw Corrosion (Nuclear FI SS System Water Service Water Strainer Pitting System Strainer Elements Corrosion Elements Inspection)

Macro-Organisms Fouling Silting The Preventive Maintenance Activities - Nuclear Service Water System Strainer Elements Inspection is a new aging management program activity. The purpose of the Preventive Maintenance Activities - Nuclear Service Water System Strainer Elements Inspection is to manage loss of material and fouling of the Nuclear Service Water System strainer elements.

This activity periodically inspects the strainer elements to verify the physical integrity is not degraded by age related degradation and fouling is not occurring. The Preventive Maintenance Activities - Nuclear Service Water System Strainer Elements Inspection is a condition monitoring program.

Scope - The scope of the Preventive Maintenance Activities - Nuclear Service Water System Strainer Elements Inspection is the strainer elements inside the McGuire Units 1 and 2 Nuclear Service Water System Strainers.

Preventive Actions - No actions are taken as part of this program to prevent aging effects or to mitigate aging degradation.

Parameters Monitored or Inspected - The Preventive Maintenance Activities - Nuclear Service Water System Strainer Elements Inspection inspects the strainer elements for signs of corrosion and fouling.

Detection of Aging Effects - In accordance with the information provided under Monitoring &

Trending below, the Preventive Maintenance Activities - Nuclear Service Water System Strainer 18.4 - 2 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 Elements Inspection will detect loss of material and fouling prior to loss of the component intended function.

Monitoring & Trending - The Preventive Maintenance Activities - Nuclear Service Water System Strainer Elements Inspection will perform direct physical inspections of the strainer elements at least once every ten years, and looks for signs of corrosion and fouling that provides evidence of age related degradation. Identification of significant defects / degradation is referred to the Corrective Action Program for further investigation.

Acceptance Criteria - The acceptance criteria for the Preventive Maintenance Activities -

Nuclear Service Water System Strainer Elements Inspection is no unacceptable loss of material or fouling of the stainless steel strainer elements that could result in the loss of component intended function as determined by engineering evaluation.

Corrective Action & Confirmation Process - Engineering evaluation is performed to disposition any significant degradation and determine continued acceptability. Specific corrective actions and confirmation are implemented in accordance with the Corrective Action Program.

Administrative Controls - Preventive Maintenance Activities - Nuclear Service Water System Strainer Elements Inspection is controlled by plant procedures and work processes. The procedures and work processes provide steps for performance of the activities and require documentation of the results.

Operating Experience - Operating experience has shown the potential for fouling of the McGuire Nuclear Station Nuclear Service Water System Strainers. Modifications have been performed to increase the reliability / efficacy of the straining / backwash function, including upgrading the strainer element material to stainless steel, providing an assured source of air to operate the backwash valve, and installation of a safety grade backwash pump / piping to ensure sufficient backwash flows. These modifications are relatively recent, hence, operating experience with the current system configuration is limited. Nonetheless, periodic physical inspection of the strainer elements is a reliable means to verify the physical condition of the strainer element and ensure that its function is not impaired.

18.4.1.2 Liquid Waste System Piping - Control Room Air Handling Units Drains The Liquid Waste System piping draining the Control Room Air Handling Units was not included in the scope of license renewal by the combined McGuire Nuclear Station, Units 1 and 2 and Catawba Nuclear Station, Units 1 and 2 License Renewal Application. It has since been determined that a break before the loop seals in this piping could allow unfiltered (possibly contaminated) air to enter the control room, and a break in the Control Building space could damage safety related electrical equipment below it. Accordingly, McGuire Nuclear Station has implemented design changes to upgrade this piping to Duke piping class F (non safety-related, seismic). This piping is now considered to perform an intended function and has been included in the scope of license renewal under 10 CFR 54.4.(a)(2).

Results from a Mechanical aging management review of the Liquid Waste System Piping for the Control Room Air Handling Units identified that this piping performs a pressure boundary intended function, and may be susceptible to the age related degradation due to loss of material. The aging management review prescribed the existing Liquid Waste System Inspection program, which has been revised to include additional inspection activities for the subject piping. A description of revisions to that program follows the aging management review summary table below.

(13 APR 2020) 18.4 - 3

UFSAR Chapter 18 McGuire Nuclear Station An interdisciplinary review of supporting features determined that there are no newly identified Civil / Structural or Electrical / I&C features associated with the Liquid Waste System Piping for the Control Room Air Handling Units.

A summary of the Mechanical aging management review for the Liquid Waste System piping draining the Control Room Air Handling Units is provided in the following table:

Internal Component Aging Component Material Environ. Aging Aging Function Management Type External Effect Mechanism Program Environ.

Crevice Corrosion Pitting Liquid Waste Raw Loss of Corrosion System Water Material Pipe PB SS Micro- Inspection biologicallly Induced Corrosion None Sheltered None None Required The Liquid Waste Inspection Program is an existing program described in Section 18.2.18 of the McGuire UFSAR. This program has been revised to include additional inspections specific to the Liquid Waste System piping draining the Control Room Air Handling Units. The following discussion reflects the program attributes described in Appendix B, Section B.3.22 of the McGuire Nuclear Station License Renewal Application, as modified to address aging management of the Liquid Waste System piping draining the Control Room Air Handling Units (new text in bold):

Scope - The scope of the Liquid Waste System Inspection is cast iron, stainless steel and carbon steel components exposed to unmonitored treated and borated water environments or raw water environments in the following McGuire systems:

  • Component Cooling System - the portion of the Component Cooling System of concern is the stainless steel waste evaporator package exposed to an unmonitored treated water environment of the Liquid Waste Recycle System.
  • Liquid Waste Recycle System - stainless steel components exposed to an unmonitored borated water environment;
  • Liquid Waste System piping draining the Control Room Air Handling Units, up to the point the piping exits the Control Building.

Preventive Actions - No actions are taken as part of this program to prevent aging effects or to mitigate aging degradation.

Parameters Monitored or Inspected - The parameters inspected by the Liquid Waste System Inspection are wall thickness, as a measure of loss of material, and visible signs of cracking and loss of material.

18.4 - 4 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 Detection of Aging Effects - The Liquid Waste System Inspection will detect the presence and extent of loss of material due to crevice and pitting corrosion and cracking due to stress corrosion / intergranular attack in stainless steel components exposed to unmonitored borated and treated water environments. In addition, this activity will detect the presence and extent of loss of material due to crevice, pitting, microbiologically influenced corrosion and cracking due to stress corrosion in stainless steel components exposed to raw water environments.

Finally, this activity will detect the presence and extent of loss of material due to crevice, general, pitting, and microbiologically influenced corrosion in carbon steel and cast iron components exposed to raw water environments.

Monitoring & Trending - The Liquid Waste System Inspection will use a volumetric technique to inspect the material/environment combinations located in each system listed above. As an alternative, visual examination will be used should access to internal surfaces become available. Selection of the specific areas for inspection for the system material/environment combinations will be the responsibility of the system engineer.

Component Cooling System At McGuire, the waste evaporator package consists of four heat exchangers. One of the four heat exchangers will be inspected. The inspection results will be applied to the other three stainless steel heat exchanger components exposed to unmonitored treated water environments.

Liquid Waste Recycle System At McGuire, the Liquid Waste System Inspection will use a combination of volumetric and visual examination of a sample population of subject components. For stainless steel components exposed to unmonitored borated water environments, the sample population will include components located in stagnant or low flow areas near collection tanks where contaminants are likely to collect and concentrate to create an environment more corrosive than the general system borated water environments. The inspection results will be applied to the stainless steel components in the unmonitored borated water environments.

Liquid Waste System Drain Lines from Control Room Air Handling Units At McGuire, the Liquid Waste System Inspection will use a combination of volumetric and visual examination at sample locations on the subject drain lines. The inspection population will include the loop seal piping where contaminants are likely to collect and concentrate to create an environment more corrosive, and other locations deemed to be representative / bounding. Given the limited extent of the piping in question, it is considered that no more than two or three inspection locations will be needed to characterize its condition. Results from localized piping inspections will be applied to the balance of piping, as appropriate.

For McGuire, this new inspection will be completed following issuance of renewed operating licenses for McGuire Nuclear Station and by June 12, 2021 (the end of the initial license of McGuire Unit 1).

No actions are taken as part of this activity to trend inspection results.

Should industry data or other evaluations indicate that the above inspections can be modified or eliminated, Duke will provide plant-specific justification to demonstrate the basis for the modification or elimination.

Acceptance Criteria - The acceptance criterion for the Liquid Waste System Inspection is no unacceptable loss of material and cracking of stainless steel components and loss of material of (13 APR 2020) 18.4 - 5

UFSAR Chapter 18 McGuire Nuclear Station carbon steel and cast iron components that could result in a loss of the component intended function(s) as determined by engineering evaluation.

Corrective Action & Confirmation Process - If engineering evaluation determines that continuation of the aging effects will not cause a loss of component intended function(s) under any current licensing basis design conditions for the period of extended operation, then no further action is required. If engineering evaluation determines that additional information is required to more fully characterize any or all of the aging effects, then additional inspections will be completed or other actions taken in order to obtain the additional information. If further engineering evaluation determines that continuation of the aging effects could cause a loss of component intended function(s) under current licensing basis design conditions for the period of extended operation, then programmatic oversight will be defined. Specific corrective actions will be implemented in accordance with the corrective action program.

Administrative Controls - The Liquid Waste System Inspection will be implemented in accordance with controlled plant procedures.

Operating Experience - The Liquid Waste System Inspection is a one-time inspection activity for which there is no operating experience.

18.4.1.3 Boral Spent Fuel Rack Neutron Attenuation Material McGuire Nuclear Station installed spent fuel racks containing Boral Neutron Attenuation Material while the combined McGuire Nuclear Station, Units 1 and 2 and Catawba Nuclear Station, Units 1 and 2 License Renewal Application was under review. However, the License Amendment to take credit for the Boral in these racks was not approved until after the renewed license was approved. As a result, the Boral Neutron Attenuation Material in these racks was not initially accorded an intended function, nor was it included in the scope of license renewal under 10 CFR 54.21(b). Since that time, McGuire Nuclear Station License Amendments 207 and 225 have been issued, allowing McGuire to take credit for the neutron attenuation capability of Boral in complying with sub-criticality margin requirements. Therefore, the Boral Neutron Attenuation Material in the McGuire Nuclear Station Spent Fuel Pool Racks is now considered to perform an intended function, and is included in the scope of license renewal under 10 CFR 54.4.(a)(1).

Results from a Civil / Structural aging management review identified that the Boral panels in the McGuire Nuclear Station spent fuel racks perform a neutron attenuation intended function, and could be susceptible to age related degradation due to reduction of neutron absorbing capacity, change in dimensions, and loss of material. A new Boral Monitoring Program has been prescribed to manage the effects of aging. A description of that program activity follows the aging management review summary table below.

An interdisciplinary review of supporting features determined that there are no newly identified Mechanical or Electrical / I&C features associated with the Boral Neutron Attenuation Material.

A summary of the Civil / Structural aging management review for the Boral Neutron Attenuation Material is provided in the following table:

18.4 - 6 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 McGuire Nuclear Station Aging Management Review Results - Other Structures 1 2 3 4 5 6 Component Component Material Environment Aging Effects Aging Type Function Management Programs and Activities Other Structural Components Boral Absorb Boral Treated Reduction of Boral Panels Neutrons Neutron- Water neutron Monitoring absorbing absorbing Program sheets capacity; change in dimensions and loss of material due to effects of SFP environment The Boral Monitoring Program is a new aging management program based on the guidance in NUREG-1801, XI.M40, Monitoring of Neutron-Absorbing Materials Other than Boraflex. The purpose of the Boral Monitoring Program is to assure that degradation of the neutron-absorbing material used in spent fuel pool storage racks that could compromise the criticality analysis will be detected. The applicable aging management program (AMP) relies on periodic inspection, testing, monitoring, and analysis of the criticality design to assure that the required sub-criticality margin is maintained during the license renewal period of extended operation.

Scope - The Boral Monitoring Program manages the effects of aging on Boral neutron-absorbing components/materials used in McGuire Nuclear Station spent fuel racks.

Preventive Actions - This Boral Monitoring Program is a condition monitoring program; therefore, there are no preventative actions.

Parameters Monitored/Inspected - Gamma irradiation and/or long-term exposure to the wet pool environment may cause loss of material and changes in dimension (such as gap formation, formation of blisters, pits and bulges) that could result in loss of neutron-absorbing capability of the material. The parameters monitored include the physical condition of the neutron-absorbing materials, such as geometric changes in the material (formation of blisters, pits, and bulges) as observed from coupons or in situ, and decreased boron areal density, etc. The parameters monitored are directly related to determination of the loss of material or loss of neutron absorption capability of the material(s).

Detection of Aging Effects - The loss of material and the degradation of the neutron absorbing material capacity are determined through coupon and / or direct in-situ testing. Such testing includes periodic verification of boron loss through areal density measurement of coupons or through direct in-situ techniques, which may include measurement of boron areal density, geometric changes in the material (blistering, pitting, and bulging), and detection of gaps through blackness testing. The frequency of the inspection and testing depends on the condition of the neutron-absorbing material and is determined considering plant specific and industry operating experience, not to exceed 10 years.

(13 APR 2020) 18.4 - 7

UFSAR Chapter 18 McGuire Nuclear Station Monitoring and Trending - The measurements from periodic inspections and analysis are compared to baseline information or prior measurements and analysis for trend analysis.

Acceptance Criteria - Although the goal is to ensure maintenance of the required sub-criticality margin for the spent fuel pool, the specific acceptance criteria for the measurements and analyses are plant specific.

Corrective Actions - Corrective actions are initiated if the results from measurements and analysis indicate that the required sub-criticality margin cannot be maintained because of current or projected future degradation of the neutron-absorbing material. Corrective actions may consist of providing additional neutron-absorbing capacity with an alternate material, or applying other options, which are available to maintain the sub-criticality margin.

Confirmation Process - Site quality assurance (QA) procedures, site review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR Part 50, Appendix B.

Administrative Controls - As discussed in the Appendix for GALL, the staff finds the requirements of 10 CFR Part 50, Appendix B, acceptable to address administrative controls.

Operating Experience - Industry operating experience reflects that degradation of neutron absorbing material is possible, with resulting decrease in neutron attenuation ability. The Boral Monitoring Program combines monitoring of coupons to verify physical condition and dimensional stability, and periodic verification of boron loss through areal density measurement of coupons or through direct in-situ techniques. These measures provide reasonable assurance that the program is able to detect degradation of the Boral neutron absorbing material in the McGuire Nuclear Station spent fuel pool prior to loss of intended function.

18.4.1.4 Earthen Dike on the North Perimeter of the McGuire Nuclear Station Site The combined McGuire Nuclear Station, Units 1 and 2 and Catawba Nuclear Station, Units 1 and 2 License Renewal Application concluded that neither the Cowans Ford Dam or the dike extension to the dam performed an intended function and therefore were not within the scope of license renewal. During a review of the McGuire response to Fukushima Near Term Task Force requirements, license basis, it was determined that the dike extension to the dam is credited with protecting safety related systems and structures from flooding from Lake Norman.

Accordingly, the Earthen dike on the north perimeter of the site is considered to perform an intended function, and is included in License Renewal scope under 10 CFR 54.4(a)(2).

Results from a Civil / Structural aging management review identified that the Earthen Dike at the McGuire Nuclear Station could be susceptible to age related degradation due to loss of material and cracking. The aging management review prescribed the existing Inspection Program for Civil Engineering Structures and Components to manage the effects of aging; this program has been revised to incorporate inspection guidance for the earth dike as described below.

An interdisciplinary review determined that there are no newly identified Mechanical or Electrical

/ I&C features associated with the Earthen Dike.

A summary of the Civil / Structural aging management review for the Earthen Dike on the north perimeter of the McGuire Nuclear Station site is provided in the following table:

18.4 - 8 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 McGuire Nuclear Station Aging Management Review Results - Other Structures 1 2 3 4 5 6 Component Component Material Environment Aging Effects Aging Type Function Management Programs and Activities Other Structural Components Earthen Flood Provides Soil External Loss of Inspection Control Dike shelter / Material Program for protection Cracking Civil to safety- Engineering related Structures and equipment Components The Inspection Program for Civil Engineering Structures and Components is an existing program described in Section 18.2.17 of the McGuire UFSAR. This program has been revised to include additional information specific to aging management of the McGuire Nuclear Station Earthen Dike. The following discussion reflects the program attributes described in Appendix B, Section B.3.21 of the McGuire Nuclear Station License Renewal Application, as modified to address aging management of the Earthen Dike (new text in bold):

Scope - The scope of the Inspection Program for Civil Engineering Structures and Components includes the following structures and the exposed external surfaces of mechanical components located within them:

McGuire Nuclear Station

  • Auxiliary Building Structures (including the Control Building, Diesel Generator Buildings, Fuel Buildings, Main Steam Doghouses)
  • Reactor Buildings (including Unit 1 and 2 internal structures, and Station Vents)
  • Standby Shutdown Facility
  • Condenser Cooling Water (RC) Intake Structure (fire pump rooms only)
  • Turbine Building (including Service Building)
  • Yard Structures (including Refueling Water Storage Tank and Reactor Make-up Water Storage Tank foundations, Refueling Water Storage Tank missile wall, trenches, and Earthen Flood Control Dike)

Preventive Actions - No actions are taken as part of this program to prevent aging effects or mitigate aging degradation.

Parameters Monitored or Inspected - The Inspection Program for Civil Engineering Structures and Components inspects the structures and the exposed external surfaces of mechanical components within them for the following:

(13 APR 2020) 18.4 - 9

UFSAR Chapter 18 McGuire Nuclear Station Concrete spalling, cracking, delaminations, honeycombs, water in-leakage, chemical leaching, peeling paint, or discoloration Masonry Walls significant cracks in joints, unsealed penetrations, missing or broken blocks, or separation from supports Structural Steel corrosion, peeling paint, beam/column deflection, loose or missing anchors/fasteners, missing or degraded grout under base plates, twisted beams, and cracked welds Equipment Foundations settlement, cracked concrete Equipment Supports cracked concrete, loose connections, corroded steel Cable Tray Supports loose connections, corrosion, distortion, and excessive deflection Roof Systems structural integrity, deteriorated penetrations (i.e., drains, vents, etc.), signs of water infiltration, cracks, ponding and flashing degradation Seismic Gaps gaps are present Siding structural integrity and visible damage Windows/Doors missing panes, cracks, deteriorated glazing, broken or cracked frames, missing or damaged hardware, and seal integrity Trenches cracks, mis-alignment or damage of covers, may spot check trenches by removing covers and inspecting walls and bottoms for cracks Earthen Structures/Dams erosion, settlement, slope stability, seepage, drainage systems, integrity of rip-rap, and environmental conditions Mechanical Components loss of material for exposed external surfaces (program will be enhanced to add this)

In addition, certain structures and structural components may be exposed to environments which make them more susceptible to degradation. Examples include, but are not limited to:

Chemical attack Sumps and chemical use areas Freeze/thaw Trench covers Excessive heat Pipe penetrations, degradation of caulking, sealants and waterstops Abrasion High traffic areas Settlement Expansion joints.

Detection of Aging Effects - In accordance with information provided in Monitoring & Trending, the Inspection Program for Civil Engineering Structures and Components will detect loss of material, cracking, and change of material properties prior to loss of structure or component intended functions.

Monitoring & Trending - Each structure or component is inspected from the interior and exterior where accessible. Some structures (or portions of structures) may be inaccessible because of radiological considerations, obstructions or other reasons. Plant specific characteristics, industry experience, and/or testing history of such structures under similar environmental conditions may be evaluated in lieu of actual inspection of the inaccessible areas. Whenever normally inaccessible areas are made accessible (i.e., by excavation or other means) an inspection is 18.4 - 10 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 performed and the results are documented as part of the Inspection Program for Civil Engineering Structures and Components. Inspections are performed by a team of at least two people. Inspectors are qualified by appropriate training and experience and approved by responsible plant management.

The Inspection Program for Civil Engineering Structures and Components is nominally performed every five years with the exact schedule being established with consideration of refueling outages for each unit. The interval may be increased to a nominal ten-year frequency with appropriate justification based on the structure, environment, and related inspection results.

The inspection will be completed in phases as necessary based on the accessibility of each structure, with the goal of completing the inspection and issuing the report within twelve months of starting the inspection. Structures are monitored in accordance with §50.65 (a)(2) provided there is no significant degradation of the structure. Structures which are determined to be unacceptable are monitored in accordance with the provisions contained in §50.65(a)(1) of the Maintenance Rule.

The Earthen Dike on the north perimeter of the site will be inspected at least once every 5 years, and will also be subject to special inspections immediately following the occurrence significant natural phenomena, such as large floods, earthquakes, hurricanes, tornadoes, and intense local rainfalls, consistent with the guidance of RG 1.127.

Trending is performed in accordance with §50.65, the Maintenance Rule. Guidance for trending per the Maintenance Rule is provided in EDM-210, Engineering Responsibilities for the Maintenance Rule, Section 210.10.

Acceptance Criteria - The acceptance criteria are no unacceptable visual indications of loss of material, cracking or change of material properties for concrete, and loss of material for steel, as identified by the accountable engineer. Acceptable structures or components are those which are capable of performing their intended function(s) until the next scheduled inspection and are considered to meet the requirements contained in §50.65(a)(2) of the Maintenance Rule.

Unacceptable structures or components are those which are damaged or degraded such that they are not capable of performing their intended function, or if degradation is to the extent and were allowed to continue uncorrected until the next normally scheduled inspection, such that the structure or component may not meet is design basis.

Corrective Actions & Confirmation Process - Structures and components not meeting the acceptance criteria are evaluated by the accountable engineer for continued service, monitoring, repair, or replacement as required. Structures and components determined to be unacceptable are required to meet the provisions contained in §50.65(a)(1) of the Maintenance Rule. Structures and components which are deemed unacceptable are documented under the corrective action program or corrected using the work management system. Specific corrective actions and confirmation actions, as needed, are implemented in accordance with the corrective action program. Subsequent inspections confirm that the corrective action was implemented and was effective.

Administrative Controls - The Inspection Program for Civil Engineering Structures and Components is implemented in accordance with a department directive.

Operating Experience - Previous inspections noted several minor degraded conditions; however, the conditions did not adversely affect the ability of the structures or components to perform their intended functions. Findings have been addressed by the corrective action program or by station work requests. Items that were noted that required additional investigation, repair or other corrective actions included: missing grout under base plates; (13 APR 2020) 18.4 - 11

UFSAR Chapter 18 McGuire Nuclear Station degraded coatings on steel, concrete, and pipe supports; minor corrosion of steel; deterioration of expansion joints; and minor cracking and spalling of concrete. Corrective actions included repair or replacement of the affected structure or structural component. The determination of specific corrective actions, including whether or not additional inspections are warranted, were made using the corrective action program.

18.4.1.5 Auxiliary Feedwater Storage Tanks McGuire Nuclear Station installed elevated Auxiliary Feedwater Storage Tanks in Units 1 & 2 while the combined McGuire Nuclear Station, Units 1 and 2 and Catawba Nuclear Station, Units 1 and 2 License Renewal Application was under review. The Auxiliary Feedwater Storage Tanks were initially installed to provide additional margin of condensate grade water for the Auxiliary Feedwater System. The Auxiliary Feedwater Storage Tanks are free standing elevated storage tanks, and are not seismically installed or provided with missile protection.

They were not initially credited with compliance with regulated events in 10 CFR 54.4(a)(3), and not included in the scope of license renewal in the combined McGuire Nuclear Station, Units 1 and 2 and Catawba Nuclear Station, Units 1 and 2 License Renewal Application. Subsequently, the Auxiliary Feedwater Storage Tanks have been credited with providing the required supply of water to the Auxiliary Feedwater System for the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> coping period under Station Blackout Conditions, as well as providing the initial supply of water for safe shutdown in the event of fire prior to switchover to long term sources. Therefore, the Auxiliary Feedwater Storage Tanks now perform an intended function, and are included in the scope of license renewal under 10 CFR 54.4(a)(3).

Results from Mechanical aging management review identified that the Auxiliary Feedwater Storage Tanks could be susceptible to age related degradation due to loss of material. The existing Inspection Program for Civil Engineering Structures and Components was prescribed to manage external surfaces of the tanks against the effects of aging. The existing Chemistry Control Program was prescribed to manage internal surfaces. Additionally, a new Preventive Maintenance Inspection Activity was prescribed to manage internal surfaces against the effects of aging. A description of that program activity follows the aging management review summary table below.

An interdisciplinary review of supporting features determined that there are also newly identified Civil / Structural features associated with the Auxiliary Feedwater Storage Tanks. Specifically, the Auxiliary Feedwater Storage Tanks are founded on piles that provide structural support, therefore perform an intended function and are also in the scope of license renewal. No newly identified Electrical / I&C features were found in association with the Auxiliary Feedwater Storage Tanks.

A summary of the Mechanical aging management review for the Auxiliary Feedwater Storage Tanks is provided in the following table:

18.4 - 12 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 McGuire Nuclear Station Auxiliary Feedwater System Component Screening and Aging Management Review Results Internal Component Component Material Environ. Aging Aging Aging Management Function Type External Effect Mechanism Program Environ.

Crevice Corrosion Chemistry Control Galvanic Program Corrosion Treated Loss of General Water Material Preventive Corrosion Maintenance Activities (Auxiliary CA Pitting Feedwater Storage Storage PB CS Corrosion Tank Internal Tanks Coating Inspection)

General Corrosion Inspection Program Loss of for Civil Material Pitting Engineering External Corrosion (Yard) Structures and Crevice Components Corrosion The Preventive Maintenance Activities - Auxiliary Feedwater Storage Tank Internal Coating Inspection is a new aging management program. The purpose of the Preventive Maintenance Activities - Auxiliary Feedwater Storage Tank Internal Coating Inspection is to manage loss of material of the internal surfaces of the carbon steel CA storage tanks. The internal carbon steel surfaces of the CA storage tanks are covered with an epoxy coating that prevents air from contacting the internal surfaces. This preventive maintenance activity inspects internal surfaces of the Auxiliary Feedwater Storage Tanks to assess the condition of the coating and to verify the absence of age related degradation. The Preventive Maintenance Activities - Auxiliary Feedwater Storage Tank Internal Coating Inspection is a condition monitoring program.

Scope - The scope of the Preventive Maintenance Activities - Auxiliary Feedwater Storage Tank Internal Coating Inspection is the internal surfaces of the McGuire Units 1 and 2 carbon steel CA storage tanks in the Auxiliary Feedwater System.

Preventive Actions - No actions are taken as part of this program to prevent aging effects or to mitigate aging degradation.

Parameters Monitored or Inspected - The Preventive Maintenance Activities - Auxiliary Feedwater Storage Tank Internal Coating Inspection inspects the internal epoxy coating for signs of blistering, chipping, peeling, and missing coating as well as signs of corrosion of the underlying carbon steel tank.

(13 APR 2020) 18.4 - 13

UFSAR Chapter 18 McGuire Nuclear Station Detection of Aging Effects - In accordance with the information provided under Monitoring &

Trending below, the Preventive Maintenance Activities - Auxiliary Feedwater Storage Tank Internal Coating Inspection will detect loss of material prior to loss of the component intended function.

Monitoring & Trending - The Preventive Maintenance Activities - Auxiliary Feedwater Storage Tank Internal Coating Inspection visually inspects the internal epoxy paint at least once every ten years. The inspection looks for signs of blistering, chipping, peeling, and missing coating as well as signs of corrosion of the underlying carbon steel tank.

Acceptance Criteria - The acceptance criteria for the Preventive Maintenance Activities -

Auxiliary Feedwater Storage Tank Internal Coating Inspection is no visual indications of coating defects that have led to corrosion of the underlying carbon steel tank surfaces.

Corrective Action & Confirmation Process - Engineering evaluation is performed to determine whether the coating and base metal continue to be acceptable. Specific corrective actions and confirmation are implemented in accordance with the corrective action program.

Administrative Controls - Preventive Maintenance Activities - Auxiliary Feedwater Storage Tank Internal Coating Inspection is controlled by plant procedures and work processes. The procedures and work processes provide steps for performance of the activities and require documentation of the results.

Operating Experience - The internal surfaces of the carbon steel refueling water storage tanks for McGuire Units 1 and 2 were inspected during outage EOC13 and EOC12, respectively, using an underwater camera. Video results showed some coating blistering, so the tanks were drained, visually inspected, and repainted in the necessary locations. No bare metal was exposed as a result of the blistering. A layer of the coating remained in the blistered location.

The submerged portion of the tanks showed little to no degradation. However, the roof, which is not a part of the pressure boundary of the tank, did show evidence of coating concerns and was blasted and repainted in several locations. This operating experience demonstrates that this activity is capable of detecting degradation prior to loss of intended function, and will be effective in managing loss of material of the carbon steel tank by maintaining the effectiveness of the epoxy coating.

Results from a Civil / Structural aging management review identified that the Auxiliary Feedwater Storage Tanks were founded on piles, which could be susceptible to age related degradation due to loss of material. The existing Inspection Program for Civil Engineering Structures and Components was prescribed to manage the effects of aging. A summary of the Civil / Structural aging management review for the Auxiliary Feedwater Storage Tanks is provided in the following table:

18.4 - 14 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 McGuire Nuclear Station Aging Management Review Results - Other Structures 1 2 3 4 5 6 Component Component Material Environment Aging Effects Aging Function Management Type Programs and Activities Other Structural Components Foundation 7 Steel Below Grade Loss of Inspection Piles (CA Material Program for Storage Tank) Civil Engineering Structures and Components Note 7 - Provides structural and/or functional support to non-safety related equipment where failure of this component could directly prevent satisfactory accomplishment of any of the required safety-related functions.

18.4.1.6 Interior Fire Protection System Turbine Building Strainers - Mechanical Elements Scoping:

A review of SSCs credited for compliance with NFPA 805 determined that additional classical fire protection features identified in the NFPA 805 Fire Safety Analysis (FSA) calculations should be included in the scope of License Renewal. Specifically, these are water based fire suppression features for the following areas in the Turbine Building of each unit:

  • Water Suppression, FA TB1 Turb Piping & Bearings
  • Water Suppression, FA TB2 Turb Piping & Bearings (13 APR 2020) 18.4 - 15

UFSAR Chapter 18 McGuire Nuclear Station Generally, a review of the expanded license renewal scope determined that the treatment of fire protection features in the License Renewal Application was bounding, such that no newly identified SSCs were involved. However, since aging management activities prescribed for fire protection strainers is component specific in the License Renewal Application (includes RAIs /

responses), this result does not apply to strainers added to scope supporting these Turbine Building fire protection features. As a result, these strainers are considered newly identified.

Unit 1 Unit 2 1 RF ST 0016 1 RF ST 0018 1 RF ST 0017 1 RF ST 0019 Resolving the expanded license renewal scope from Fire Safety Area evaluations determined that the following Turbine Building Fire Protection strainers are newly identified SSCs:

Screening:

The Turbine Building Fire Protection System strainers provide flow to credited fire protection features, as well as filtration to ensure that foreign material does not adversely impact system intended functions. As such, the strainer housings have a pressure boundary intended function whereas the strainer baskets have a filtration intended function. Both the strainer housings and baskets perform their intended functions in a passive manner, and are therefore subject to aging management review.

Aging Management Review:

A review of vendor information reflects that the subject strainers are constructed with a carbon steel housing and stainless steel or monel basket. The subject strainers are installed in the Turbine Building, and the strainer housings are considered to be exposed to a sheltered external environment. The Fire Protection System is charged with lake water. The internal environment is considered to be lake water, and has the potential to accumulate contaminants resulting in fouling.

Aging management review of the Turbine Building Fire Protection Strainers is summarized in the following Aging Management Review (AMR) table:

18.4 - 16 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 Component Component Material Internal Aging Effect Aging Type Function Environment Management Program and External Activities Environment Turbine Pressure Carbon Steel Raw Water Loss of Service Building Fire Boundary Material Water Piping Protection Corrosion Strainer Program Housings Sheltered Loss of Inspection Material Program For Civil Engineering Structures and Components Turbine Filtration Stainless Raw Water 1 Loss of Turbine Building Fire Steel / Monel Material Building Fire Protection Protection Strainer Strainer Baskets Inspections Fouling Turbine Building Fire Protection Strainer Inspections No External Not Not Environment Applicable Applicable Note (1) - Raw water considered to constitute both internal and external surface environments for er baskets Aging Management Programs:

Service Water Piping Corrosion Program: UFSAR Section 18.2.25 provides the following description of the scope of the Service Water Piping Corrosion Program:

Scope - For license renewal, the Service Water Piping Corrosion Program is credited with managing loss of material for components in the following systems:

  • Containment Ventilation Cooling Water
  • Exterior Fire Protection
  • Condenser Circulating Water
  • Interior Fire Protection (13 APR 2020) 18.4 - 17

UFSAR Chapter 18 McGuire Nuclear Station

  • Nuclear Service Water The UFSAR description further states that the Service Water Piping Corrosion Program uses carbon steel piping components exposed to raw water as a leading indicator of the general material condition of system components. As described in the Monitoring and Trending Program element:

The intent of the Service Water Piping Corrosion Program is to inspect a number of locations with conditions that are characteristic of the conditions found throughout the raw water systems above. The results of these inspection locations would then be applied to similar locations throughout all the raw water systems within the scope of license renewal.

This characteristic-based approach recognizes the commonality among the component materials of construction and the environment to which they are exposed. Inspection results are used to determine and expand, as necessary, the number of inspection locations in a given characteristic set.

Since the program does not identify specific monitoring locations and the systems addressed by the program already includes the Interior Fire Protection System, adding the Turbine Building Fire Protection Strainer Housings to the scope of the program does not impact the program description or applicability. Hence, no additional actions are needed to address the use of this program to manage loss of material of internal surfaces of the steel strainer housings.

Inspection Program For Civil Engineering Structures and Components: UFSAR Section 18.2.17 contains the following information regarding the Inspection Program For Civil Engineering Structures and Components:

The Inspection Program for Civil Engineering Structures and Components is intended to meet the requirements of 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants (the Maintenance Rule). This program:

  • monitors and assesses mechanical components, civil structures and components and their condition in order to provide reasonable assurance that they are capable of performing their intended functions in accordance with the current licensing basis;
  • includes nuclear safety-related structures which enclose, support, or protect nuclear safety-related systems and components, non-safety-related structures whose failure may prevent a nuclear safety-related system or component from fulfilling its intended function, and non safety-related structures which support equipment relied on during certain regulated events.

As described, the Inspection Program For Civil Engineering Structures and Components uses periodic monitoring to identify evidence of age related degradation. The program includes a specific commitment to incorporate monitoring of exposed external surfaces of mechanical components, which bounds the inclusion of the Turbine Building Fire Protection System Strainers. No additional inspections or activities are deemed necessary to credit this program.

Turbine Building Fire Protection Strainer Inspections: License Renewal commitments include a set of enhancements added to the Fire Protection Program to address specific aging management issues identified during the review of the License Renewal Application. These commitments are individually identified and described in Section 18.2.8 of the UFSAR, and include specific activities relating to the Main Fire Pump Strainers and the Jockey Pump 18.4 - 18 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 Strainers. To address aging management of the newly identified Turbine Building Fire Protection System Strainers, a sixth activity, Turbine Building Fire Protection Strainer Inspections, will be added. The following is a ten element description of the new Turbine Building Fire Protection Strainer Inspections activity:

Scope - The scope of the Turbine Building Fire Protection Strainer Inspections is the following set of in line strainers located in the Turbine Building:

Unit 1 Unit 2 1 RF ST 0016 1 RF ST 0018 1 RF ST 0017 1 RF ST 0019 Preventive Actions - No actions are taken as part of this program to prevent aging effects or to mitigate aging degradation.

Parameters Monitored or Inspected - The parameters inspected by the Turbine Building Fire Protection Strainer Inspections are loss of material and fouling due to exposure to a raw water environment.

Detection of Aging Effects - In accordance with information provided in Monitoring & Trending below, the Turbine Building Fire Protection Strainer Inspections will detect loss of material and fouling of the Turbine Building Fire Protection Strainers prior to loss of component intended function.

Monitoring & Trending - The Turbine Building Fire Protection Strainer Inspections is a general visual inspection for loss of material and fouling of the strainer baskets. This inspection will be performed at a frequency of at least once every 10 years.

Acceptance Criteria - The acceptance criteria for the Turbine Building Fire Protection Strainer Inspections is no unacceptable loss of material or fouling that could result in a loss of component intended function(s) as determined by engineering.

Corrective Action & Confirmation Process - If engineering evaluation determines that the observed aging effects do not cause a loss of component intended function, then no further actions are necessary. If engineering evaluation determines that the observed aging effects could cause a loss of component intended function, then corrective actions are taken, including cleaning of the strainer or replacement. Specific corrective actions will be implemented in accordance with the corrective action program.

Administrative Controls - The Turbine Building Fire Protection Strainer Inspections will be implemented in accordance with controlled plant procedures.

Operating Experience - The Turbine Building Fire Protection Strainer Inspections is a newly credited license renewal inspection. Visual inspection is an effective method for detecting age-related degradation in the strainers. These strainer baskets are been inspected and cleaned periodically, and these activities have been effective in managing fouling and loss of material.

The Turbine Building Fire Protection Strainer Inspections activity will provide reasonable assurance that loss of material and/or fouling will not impact the strainer basket intended function during the period of extended operation.

(13 APR 2020) 18.4 - 19

UFSAR Chapter 18 McGuire Nuclear Station 18.4.1.7 Main Condenser Hotwell and Condensate System Piping Scoping:

The Main Condenser Hotwell is in the scope of license renewal as a suction source for the Auxiliary Feedwater Pumps, to support decay heat removal in the event of a fire under certain Non-Power Operation (NPO) scenarios. Connected piping between the Main Condenser Hotwell and the Turbine Driven Auxiliary Feedwater Pump suction piping also performs an intended function and is added to the scope of license renewal.

Screening:

While the hotwell is open to and directly communicates with the balance of the condenser, only the portion of the hotwell which collects and stores water is in the scope of license renewal. The Condensate System piping components on the flow path to the Turbine Driven Auxiliary Feedwater Pump suction is also in the scope of license renewal. Both the Main Condenser Hotwell and Condensate System Piping are passive SSCs, and perform a pressure boundary intended function.

Aging Management Review:

The Main Condenser Hotwell is a steel structure, located directly beneath the main condenser, and penetrated by numerous piping penetrations. During normal operations, the hotwell collects condensed steam for reprocessing in the steam cycle. As such, the internal environment is considered to be treated water. The main condenser normally operates under a significant vacuum, and the ability of the condenser to maintain vacuum is integral to the operation of the main turbine, and to power production itself. As such, the integrity of the main condenser and hotwell is continually monitored during power operations by the measure of condenser vacuum.

The Condensate System Piping in the scope of license renewal is constructed of carbon steel.

Water in the Main Condenser Hotwell and Condensate System Piping is purified, chemically treated, demineralized water.

The following aging effects are considered to be applicable to internal surfaces of the Main Condenser Hotwell and Condensate System Piping:

  • Loss of Material due to General Corrosion
  • Loss of Material due to Pitting Corrosion
  • Loss of Material due to Crevice Corrosion The Chemistry Control Program is credited with aging management of internal surfaces of the Main Condenser Hotwell and Condensate System Piping. Augmented physical inspections of internal surfaces of the hotwell and Condensate System Piping are not deemed necessary to ensure the intended function is maintained due to (1) the continuous monitoring of condenser integrity during power operations, and (2) inspections of Main Feedwater and Auxiliary Feedwater components already prescribed by the MNS License Renewal Application to ensure the effectiveness of secondary system chemistry controls.

Application of the AMR logic for external surfaces of Carbon Steel in a sheltered environment reflects that the aging effect of loss of material due to General Corrosion is applicable.

Consistent with the methodology in the License Renewal Application, aging management of external surfaces of the Main Condenser Hotwell and Condensate System Piping is accomplished by the Inspection Program for Civil Engineering Structures and Components, under the existing commitment to perform inspections of exposed external surfaces of 18.4 - 20 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 mechanical components. Aging management review of the Main Condenser Hotwell and Condensate System Piping is summarized in the following AMR table:

Component Component Material Internal Aging Aging Management Type Function Environment Effect Program and Activities External Environment Main PB CS Treated Loss of Chemistry Control Condenser Water Material Program Hotwell Sheltered Loss of Inspection Program Material For Civil Engineering Structures and Components Condensate PB CS Treated Loss of Chemistry Control System Piping Water Material Program Sheltered Loss of Inspection Material Program For Civil Engineering Structures and Components Aging Management Programs:

The Chemistry Control Program is credited with aging management of internal surfaces of the Main Condenser Hotwell and Condensate System Piping. UFSAR Section 18.2.4 has the following description of the Chemistry Control Program The purpose of the Chemistry Control Program is to manage loss of material and/or cracking of components exposed to borated water, closed cooling water, fuel oil, and treated water environments. This program manages the relevant conditions that lead to the onset and propagation of loss of material, cracking, and fouling which could lead to a loss of structure or component intended functions. Relevant conditions are specific parameters such as halogens, dissolved oxygen, conductivity, biological activity, and corrosion inhibitor concentrations that could lead to loss of material and/or cracking if not properly controlled.

The Chemistry Control Program contains system specific acceptance criteria that are based on the guidance provided in EPRI PWR Primary Water Chemistry Guidelines, EPRI PWR Secondary Water Chemistry Guidelines, and EPRI Closed Cooling Water Chemistry Guideline.

The Chemistry Control Program is generically applicable to components in secondary side, treated water applications, and the program is not impacted by the inclusion of the Main Condenser Hotwell and Condensate System Piping. No additional inspections or activities are deemed necessary.

The Inspection Program For Civil Engineering Structures and Components will be used to manage aging effects of external surfaces of the Main Condenser Hotwell and Condensate (13 APR 2020) 18.4 - 21

UFSAR Chapter 18 McGuire Nuclear Station System Piping. UFSAR Section 18.2.17 contains the following information regarding the Inspection Program For Civil Engineering Structures and Components:

The Inspection Program for Civil Engineering Structures and Components is intended to meet the requirements of 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants (the Maintenance Rule). This program:

(1) monitors and assesses mechanical components, civil structures and components and their condition in order to provide reasonable assurance that they are capable of performing their intended functions in accordance with the current licensing basis; (2) includes nuclear safety-related structures which enclose, support, or protect nuclear safety-related systems and components, non-safety-related structures whose failure may prevent a nuclear safety-related system or component from fulfilling its intended function, and non safety-related structures which support equipment relied on during certain regulated events.

As described, this program uses periodic monitoring to identify evidence of age related degradation. The program includes a specific commitment to incorporate monitoring of exposed external surfaces of mechanical components. No additional inspections or activities are deemed necessary to credit this program.

18.4.1.8 Upper Surge Tanks and Condensate Storage System Piping Scoping:

Each units Condensate Storage System is provided with two Upper Surge Tanks. The Upper Surge Tanks are in the scope of license renewal as a suction source for the Auxiliary Feedwater Pumps, to support decay heat removal in the event of a fire under certain Non-Power Operation (NPO) scenarios. Connected piping between the Upper Surge Tanks and the Auxiliary Feedwater Pumps is also in the scope of license renewal, but is addressed by existing line items in the Condensate Storage System and Auxiliary Feedwater System.

Screening:

The Upper Surge Tanks and Condensate Storage System Piping on the flow path to the Auxiliary Feedwater Pumps are a passive components, and subject to aging management review. These tanks have a pressure boundary intended function.

Aging Management Review:

Based on outline drawings, the Upper Surge Tanks are constructed of stainless steel. The Condensate Storage System piping flow path from the USTs to the Auxiliary Feedwater System suction piping to the Turbine Driven Auxiliary Feedwater Pump is constructed of carbon steel.

During normal operations, the Upper Surge Tanks store condensate for makeup to the Main Condenser Hotwell, and also provide a source of condensate grade water for the Auxiliary Feedwater Pumps. As such, the internal environment of the Upper Surge Tanks and Condensate Storage System Piping is considered to be treated water.

The following aging effects are considered to be applicable to internal surfaces of the Upper Surge Tanks:

  • Loss of Material due to Pitting Corrosion
  • Loss of Material due to Crevice Corrosion 18.4 - 22 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 The Chemistry Control Program is credited with aging management of internal surfaces of the Upper Surge Tanks. Physical inspections of internal surfaces of these tanks are not deemed necessary considering (1) construction of corrosion resistant material, and (2) confirmatory inspections of Chemistry Control Program effectiveness are already required in the Main Feedwater and Auxiliary Feedwater Systems by the License Renewal Application (Ref.

discussion of Open Item 3.4.1.2.2-1 in NUREG-1772).

The following aging effects are considered to be applicable to internal surfaces of Condensate Storage System piping / piping components:

  • Loss of Material due to General Corrosion
  • Loss of Material due to Pitting Corrosion
  • Loss of Material due to Crevice Corrosion The Upper Surge Tanks and Condensate Storage System Piping are located within the Turbine Buildings, and considered to be in a sheltered environment. Application of the AMR logic for Stainless Steel in a sheltered environment reflects that no aging effects are considered applicable to external surfaces of the stainless steel Upper Surge Tanks. The carbon steel Condensate Storage System Piping is susceptible to general corrosion, which is managed by the Inspection Program For Civil Engineering Structures and Components .

Aging management review of the Upper Surge Tanks and Condensate Storage System Piping is summarized in the following AMR table:

Component Component Material Internal Aging Aging Management Type Function Environment Effect Program and Activities External Environment Upper Surge PB SS Treated Loss of Chemistry Control Tanks Water Material Program Sheltered None None Required Identified Condensate PB CS Treated Loss of Chemistry Control Storage Water Material Program System piping

/ piping components Sheltered Loss of Inspection Program Material For Civil Engineering Structures and Components (13 APR 2020) 18.4 - 23

UFSAR Chapter 18 McGuire Nuclear Station Aging Management Programs:

The Chemistry Control Program is credited with aging management of internal surfaces of the Upper Surge Tanks and Condensate Storage System Piping. UFSAR Section 18.2.4 has the following description of the Chemistry Control Program The purpose of the Chemistry Control Program is to manage loss of material and/or cracking of components exposed to borated water, closed cooling water, fuel oil, and treated water environments. This program manages the relevant conditions that lead to the onset and propagation of loss of material, cracking, and fouling which could lead to a loss of structure or component intended functions. Relevant conditions are specific parameters such as halogens, dissolved oxygen, conductivity, biological activity, and corrosion inhibitor concentrations that could lead to loss of material and/or cracking if not properly controlled.

The Chemistry Control Program contains system specific acceptance criteria that are based on the guidance provided in EPRI PWR Primary Water Chemistry Guidelines, EPRI PWR Secondary Water Chemistry Guidelines, and EPRI Closed Cooling Water ChemistryGuideline.

The Chemistry Control Program is generically applicable to components in secondary side, treated water applications, and the program is not impacted by the inclusion of the Upper Surge Tanks and Condensate Storage System Piping. No additional inspections or activities are deemed necessary.

The Inspection Program For Civil Engineering Structures and Components will be used to manage aging effects of external surfaces of the Upper Surge Tanks. UFSAR Section 18.2.17 contains the following information regarding the Inspection Program For Civil Engineering Structures and Components:

The Inspection Program for Civil Engineering Structures and Components is intended to meet the requirements of 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants (the Maintenance Rule). This program:

(1) monitors and assesses mechanical components, civil structures and components and their condition in order to provide reasonable assurance that they are capable of performing their intended functions in accordance with the current licensing basis; (2) includes nuclear safety-related structures which enclose, support, or protect nuclear safety-related systems and components, non-safety-related structures whose failure may prevent a nuclear safety-related system or component from fulfilling its intended function, and non safety-related structures which support equipment relied on during certain regulated events.

As described, this program uses periodic monitoring to identify evidence of age related degradation. The program includes a specific commitment to incorporate monitoring of exposed external surfaces of mechanical components. No additional inspections or activities are deemed necessary to credit this program.

18.4.1.9 Reactor Makeup Water Storage Tanks and Reactor Makeup Water Pumps Scoping:

Each unit's Boron Recycle System is provided with a Reactor Makeup Water Storage Tank. The Reactor Makeup Water Storage Tank receives distillate from the recycle evaporator for use as makeup to the Reactor Coolant System and flush water for components within the Auxiliary and Reactor Buildings. The RMWSTs are in the scope of license renewal as an inventory makeup 18.4 - 24 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 source for the spent fuel pools. This is not a normal function of the RMWSTs, but is credited in NFPA 805 Non-Power Evaluations to provide inventory and support decay heat removal from the spent fuel pools in the event of a fire under certain Non-Power Operation (NPO) scenarios.

Connected piping / piping components between the Reactor Makeup Water Storage Tanks and the spent fuel pools is also in the scope of license renewal, but are generally addressed by existing stainless steel line License Renewal Application AMR items in the Boron Recycle System and the Spent Fuel Cooling System. The Reactor Makeup Water Pumps are on this flowpath, and will be uniquely identified for the purpose of aging management review, consistent with the methodology in the License Renewal Application. Therefore, the RMWSTs and Reactor Makeup Water Pumps are considered newly identified for the purposes of this assessment.

Screening:

The RMWSTs are passive components, and subject to aging management review. For the Reactor Makeup Water Pumps, only the casings are passive and subject to aging management review. These RMWSTs and Reactor Makeup Water Pumps have a pressure boundary intended function.

Aging Management Review:

Based on outline drawings, the RMWSTs are constructed of carbon steel with stainless steel nozzles. They are coated internally with an phenolic coating, and fitted with a polymeric diaphragm which is attached at the top of the tank and drapes over the contents of the tank. The purpose of the diaphragm is to limit the oxygenation of the tank water in support of improved water chemistry. The diaphragm does not perform an intended function, and is not subject to aging management review. The Reactor Makeup Water Pumps are constructed of stainless steel.

During normal operations, the RMWST provide a source of primary grade water for the makeup to the Reactor Coolant (NC) System. As such, the internal environment of the RMWSTs and Reactor Makeup Water Pumps is considered to be treated water.

The following aging effects are considered to be applicable to internal surfaces of the RMWSTs and Reactor Makeup Water Pumps:

  • Loss of Material due to Pitting Corrosion
  • Loss of Material due to Crevice Corrosion
  • Loss of Material due to General Corrosion (RMWST only)

The Preventive Maintenance Activities - Reactor Makeup Water Storage Tank Internal Coating Inspection is credited with aging management of internal surfaces of the RMWSTs. This activity is closely aligned with the Refueling Water Storage Tank Internal Coating Inspection defined in the License Renewal Application, which is credited with aging management of internal surfaces of the Refueling Water Storage Tanks (FWSTs).

The FWSTs and RMWSTs are similar from a design and functional standpoint, and sit immediately adjacent to each other on either side of the Reactor Containment Buildings.

The RMWSTs are located in the yard, and considered to be in an outdoor environment.

Application of the AMR logic for carbon steel in an outdoor (Yard) environment reflects that loss of material is considered applicable to external surfaces. The stainless steel Reactor Makeup Water Pumps are located in the Auxiliary Building, in a sheltered environment. No aging effects are predicted for external surfaces of these components.

Aging management review of the RMWSTs is summarized in the following AMR table:

(13 APR 2020) 18.4 - 25

UFSAR Chapter 18 McGuire Nuclear Station Component Component Material Internal Aging Aging Management Type Function Environment Effect Program and Activities External Environment Reactor PB CS Treated Loss of Preventive Makeup Water Material Maintenance Water Activities -

Storage Reactor Makeup Tank Water Storage Tanks Internal Coating Inspection Yard Loss of Inspection Material Program for Civil Engineering Structures and Components SS Treated Loss of Chemistry Control Water Material Program Yard None None Required Reactor PB SS Treated Loss of Chemistry Control Makeup Water Material Program Water Pumps (casing only)

Sheltered None None Required Aging Management Programs:

The Chemistry Control Program is credited with aging management of internal surfaces of the RMWST and Reactor Makeup Water Pumps. UFSAR Section 18.2.4 has the following description of the Chemistry Control Program The purpose of the Chemistry Control Program is to manage loss of material and/or cracking of components exposed to borated water, closed cooling water, fuel oil, and treated water environments. This program manages the relevant conditions that lead to the onset and propagation of loss of material, cracking, and fouling which could lead to a loss of structure or component intended functions. Relevant conditions are specific parameters such as halogens, 18.4 - 26 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 dissolved oxygen, conductivity, biological activity, and corrosion inhibitor concentrations that could lead to loss of material and/or cracking if not properly controlled.

The Chemistry Control Program contains system specific acceptance criteria that are based on the guidance provided in EPRI PWR Primary Water Chemistry Guidelines, EPRI PWR Secondary Water Chemistry Guidelines, and EPRI Closed Cooling Water Chemistry Guideline.

The Chemistry Control Program is generically applicable to components in secondary side, treated water applications, and the program is not impacted by the inclusion of the RMWST and Reactor Makeup Water Pumps. No additional inspections or activities are deemed necessary.

The Inspection Program For Civil Engineering Structures and Components will be used to manage aging effects of external surfaces of the RMWSTs. UFSAR Section 18.2.17 contains the following information regarding the Inspection Program For Civil Engineering Structures and Components:

The Inspection Program for Civil Engineering Structures and Components is intended to meet the requirements of 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants (the Maintenance Rule). This program:

(1) monitors and assesses mechanical components, civil structures and components and their condition in order to provide reasonable assurance that they are capable of performing their intended functions in accordance with the current licensing basis; (2) includes nuclear safety-related structures which enclose, support, or protect nuclear safety-related systems and components, non-safety-related structures whose failure may prevent a nuclear safety-related system or component from fulfilling its intended function, and non safety-related structures which support equipment relied on during certain regulated events.

As described, this program uses periodic monitoring to identify evidence of age related degradation. The program includes a specific commitment to incorporate monitoring of exposed external surfaces of mechanical components. No additional inspections or activities are deemed necessary to credit this program.

The Preventive Maintenance Activities - Reactor Makeup Water Storage Tank Internal Coating Inspection is credited with aging management of internal (coated) surfaces of the RMWSTs.

This program is closely aligned with the Refueling Water Storage Tank Internal Coating Inspection, which was credited in the License Renewal Application with aging management of the Refueling Water Storage Tank. The following description ten element is provided for the Reactor Makeup Water Storage Tank Internal Coating Inspection The purpose of the Preventive Maintenance Activities - Reactor Makeup Water Storage Tank Internal Coating Inspection is to manage loss of material of the internal surfaces of the carbon steel refueling water storage tanks. The internal carbon steel surfaces of the refueling water storage tank are coated with a phenolic epoxy paint that prevents borated water and air from contacting the internal surfaces. Continued presence of an intact coating precludes loss of material of the internal surfaces of the carbon steel refueling water storage tank that could lead to loss of pressure boundary function. This preventive maintenance activity inspects the internal coating of the refueling water storage tanks to check the condition of the coating and to identify coating failures. The Preventive Maintenance Activities - Reactor Makeup Water Storage Tank Internal Coating Inspection is a condition monitoring program.

(13 APR 2020) 18.4 - 27

UFSAR Chapter 18 McGuire Nuclear Station Scope - The scope of the Preventive Maintenance Activities - Reactor Makeup Water Storage Tank Internal Coating Inspection is the internal surface of the McGuire Units 1 and 2 carbon steel Reactor Makeup Water Storage Tanks (RMWSTs) in the Boron Recycle System.

Preventive Actions - No actions are taken as part of this program to prevent aging effects or to mitigate aging degradation.

Parameters Monitored or Inspected - The Preventive Maintenance Activities - Reactor Makeup Water Storage Tank Internal Coating Inspection inspects the phenolic epoxy paint for signs of blistering, chipping, peeling, and missing paint as well as signs of corrosion of the underlying carbon steel tank.

Detection of Aging Effects - In accordance with the information provided under Monitoring &

Trending below, the Preventive Maintenance Activities - Reactor Makeup Water Storage Tank Internal Coating Inspection will detect loss of material prior to loss of the component intended function.

Monitoring & Trending - The Preventive Maintenance Activities - Reactor Makeup Water Storage Tank Internal Coating Inspection visually inspects the internal phenolic epoxy paint every ten years using an underwater video camera. The inspection looks for signs of blistering, chipping, peeling, and missing paint as well as signs of corrosion of the underlying carbon steel tank. Detection of defects in the internal coating results in draining of the tank for further inspection and evaluation of the defects.

Acceptance Criteria - The acceptance criteria for the Preventive Maintenance Activities -

Reactor Makeup Water Storage Tank Internal Coating Inspection is no visual indications of coating defects that have led to corrosion of the underlying carbon steel tank surfaces.

Corrective Action & Confirmation Process - Engineering evaluation is performed to determine whether the coating and base metal continue to be acceptable. Specific corrective actions and confirmation are implemented in accordance with the corrective action program.

Administrative Controls - Preventive Maintenance Activities - Reactor Makeup Water Storage Tank Internal Coating Inspection is controlled by plant procedures and work processes. The procedures and work processes provide steps for performance of the activities and require documentation of the results.

Operating Experience - The internal surfaces of the Reactor Makeup Water Storage Tanks for McGuire Units 1 and 2 are currently inspected on a three year frequency. Photographic documentation of these inspections show little degradation. This operating experience demonstrates that this activity when continued through the extended period of operation will continue to be effective in managing loss of material of the carbon steel tank by maintaining the effectiveness of the phenolic epoxy paint.

Conclusion The Preventive Maintenance Activities - Reactor Makeup Water Storage Tank Internal Coating Inspection has been demonstrated to be capable of managing loss of material of the internal carbon steel by maintaining an intact protective coating. The Preventive Maintenance Activities - Reactor Makeup Water Storage Tank Internal Coating Inspection described above is similar to the Preventive Maintenance Activities - Refueling Water Storage Tank Internal Coating Inspection, which is credited in the License Renewal Application with aging management of the Refueling Water Storage Tank. Based upon the above review, the continued implementation of the Preventive Maintenance Activities -

18.4 - 28 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 Reactor Makeup Water Storage Tank Internal Coating Inspection provides reasonable assurance that the aging effects will be managed and that the component will continue to perform its intended function(s) for the period of extended operation.

18.4.1.10 Radwaste Facility Scope: Fire Protection Radiological Release evaluations performed in support of the transition to NFPA 805 reflect that the Radwaste Facility is a hardened structure that provides a containment to limit radioactive releases resulting from fire fighting activities. The features of the building credited in this regard are the hardened walls in the building that limit the spread of the fire and contain the products of fire fighting. Piping systems within the building, including the Waste Evaporator Tank and Floor Drain Tank, are not credited with an intended function in this regard. The Radwaste Facility houses ventilation system equipment that is credited with monitoring releases from the building, and isolating if necessary to limit releases.

Screening:

For the purpose of screening, the Radwaste Facility is included in the categorization of Other Structures, as addressed in the Section 3.5-2 of the License Renewal Application. Since the only intended functions of this building relate to containment of radiological releases from fire fighting activities, only those structural elements that participate in this function are in the scope of license renewal. Specifically, this pertains to the floor, walls, roof, and exterior structural elements. The Radwaste Facility also houses ventilation system equipment that is credited with monitoring releases from the building, and isolating if necessary to limit releases.

Accordingly, the intended functions ascribed to the Radwaste Facility are to provide a pressure boundary and/or fission product barrier and structural support and/or shelter to components relied on during certain postulated fire, anticipated transients without scram, and/or station blackout events.

Aging Management Review:

Aging management review of the Radwaste Facility was accomplished consistent with the methodology in the License Renewal Application, as modified by the revised AMR Table 3.5-2 provided in the letter dated October 28, 2002 (ADAMS Accession No. ML023090324). The Radwaste Facility is constructed of reinforced concrete around the tanks and process areas.

Consistent with the methodology in the License Renewal Application, the internal environments is considered to be sheltered, and the external environment is considered to be External (exposed to weather). Portions of the foundation are below grade. While it is possible for the Radwaste Facility tanks to contain borated fluid, it is considered unlikely that leakage from the tank rooms could occur that would pose a liability to the structure itself or to the portion of the ventilation system that is in the scope of license renewal, and boric acid wastage is not considered applicable.

Based on the review of Radwaste Facility design details and consistent with License Renewal Application AMR Table 3.5-2 as modified by letter dated October 28, 2002, applicable aging effects for the Radwaste Facility are loss of material and cracking. The Inspection Program for Civil Engineering Structures and Components is credited with managing these aging effects.

Aging management review for equipment and component supports within the scope of license renewal for the Radwaste Facility are likewise bounded by the AMR line items identified in License Renewal Application AMR Table 3.5-3, and also credit the Inspection Program for Civil Engineering Structures and Components for aging management of loss of material.

(13 APR 2020) 18.4 - 29

UFSAR Chapter 18 McGuire Nuclear Station Aging Management Programs:

The Inspection Program For Civil Engineering Structures and Components was an existing program in the License Renewal Application (Section B.3.21). UFSAR Section 18.2.17 contains the following description of the Inspection Program For Civil Engineering Structures and Components:

The Inspection Program for Civil Engineering Structures and Components is intended to meet the requirements of 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants (the Maintenance Rule). This program:

(1) monitors and assesses mechanical components, civil structures and components and their condition in order to provide reasonable assurance that they are capable of performing their intended functions in accordance with the current licensing basis; (2) includes nuclear safety-related structures which enclose, support, or protect nuclear safety-related systems and components, non-safety-related structures whose failure may prevent a nuclear safety-related system or component from fulfilling its intended function, and non safety-related structures which support equipment relied on during certain regulated events.

NEI 96-03, Industry Guideline for Monitoring the Condition of Structures at Nuclear Power Plants, has been used as guidance in the preparation of the Inspection Program for Civil Engineering Structures and Components. Examination and assessment of the condition of a structure is performed using guidance provided in codes and standards such as:

  • ACI 349.3, Evaluation of Existing Nuclear Safety-Related Concrete Structures Specific corrective actions are implemented in accordance with the Problem Investigation Process. The Problem Investigation Process applies to all structures and components within the scope of the Inspection Program for Civil Engineering Structures and Components.

Specification DPS-1274.00-00-0005, License Renewal Aging Management Programs and Activities, provides the basis for MNS aging management activities, updated through the receipt of NUREG-1772, License Renewal Safety Evaluation Report (SER). This specification contains a 10 element description of aging management programs, including the Inspection Program For Civil Engineering Structures and Components. A review of the 10 element program description DPS-1274.00-00-0005 finds the following changes (in bold) are needed to address inspection of the Radwaste Facility Program

Description:

The Inspection Program for Civil Engineering Structures and Components is credited with managing the following aging effects for the period of extended operation:

  • Loss of material due to corrosion for exposed surfaces of steel components:

anchorage / embedments; cable tray and conduit supports; checkered plates; equipment component supports; expansion anchors; flood, pressure, and specialty doors; HVAC duct supports; instrument racks and frames; lead shielding supports; metal siding; pipe supports; stair, platform, and grating supports; structural steel beams, columns, plates and trusses; sump screens, Shield Building penetrations; and the unit vent stack

  • Cracking of masonry block walls 18.4 - 30 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18

  • Change in material properties due to leaching of concrete walls and roofs
  • Loss of material and cracking for reinforced concrete beams, columns, and walls for the Nuclear Service Water and Standby Nuclear Service Water Pump Structure and the Low Pressure Service Water Intake Structure (Catawba only)
  • Cracking and change in material properties of elastomeric flood seals (Catawba only)
  • Loss of material of composite roofing
  • Loss of material of exposed external surfaces of mechanical components (program will be enhanced to add this)
  • Loss of material of the steel components of the Yard Drainage System (program will be enhanced to add this for Catawba only)
  • Although Duke did not identify any aging effects for other concrete components beyond those identified above, the NRC staff in its SER dated August 14, 2002 identified loss of material, cracking and change in material property to be both plausible and applicable aging effects for all concrete components.

Notwithstanding the disagreement on the aging effects that require management for the period of extended operation, Duke committed, in its response to Open Items 3.5-1 and 3.5-3 provided in letters dated October 2, 2002 and November 14, 2002, to perform periodic inspections of the remaining concrete components to manage the aging effects of loss of material, cracking, and change in material properties using the Inspection Program for Civil Engineering Structures and Components.

The Inspection Program for Civil Engineering Structures and Components is applicable in meeting the regulatory requirements of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants. The Inspection Program for Civil Engineering Structures and Components is a condition monitoring program.

Additional Information: The Maintenance Rule requires that the performance condition of structures be monitored in a manner sufficient to provide reasonable assurance that those structures are capable of fulfilling their intended function. Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, NUMARC 93-01 [Reference 1],

paragraph 9.4.2.4 provides examples of structural monitoring activities including visual inspection and non-destructive examination. NUREG-1526 [Reference 2] states that certain structures such as the primary containment can be monitored through the performance of established testing requirements however, other structures such as reactor buildings, auxiliary buildings, and cooling towers, may be more amenable to condition monitoring. Structures subject to this examination are listed in EDM-410, Inspection Program for Civil Engineering Structures and Components, Table 410-2 [Reference 3]. The inspection program contains more than the structures within license renewal scope because it addresses structures within the scope of the Maintenance Rule.

The Inspection Program for Civil Engineering Structures and Components is evaluated for each program element in accordance with the guidance provided in Appendix C of specification DPS-1274.00-00-0001, License Renewal Technical Information Development Guide [Reference 4] to provide reasonable assurance that the effects of aging will be adequately managed for the components so that their intended function(s) will be maintained consistent with the CLB for the period of extended operation. The evaluation results are provided below.

(13 APR 2020) 18.4 - 31

UFSAR Chapter 18 McGuire Nuclear Station Scope - The scope of the Inspection Program for Civil Engineering Structures and Components includes the following structures and the exposed external surfaces of mechanical components located within them:

McGuire Nuclear Station

  • Auxiliary Building Structures (including the Control Building, Diesel Generator Buildings, Fuel Buildings, Main Steam Doghouses)
  • Reactor Buildings (including Unit 1 and 2 internal structures, and Station Vents)
  • Standby Shutdown Facility
  • Condenser Cooling Water (RC) Intake Structure (fire pump rooms only)
  • Turbine Building (including Service Building)
  • Yard Structures (including Refueling Water Storage Tank and Reactor Make-up Water Storage Tank foundations, Refueling Water Storage Tank missile wall, Transformer Station and Switchyard electrical enclosures and equipment pads, Relay House, trenches (pipe and cable), equipment and component supports located outside)
  • Radwaste Facility Monitoring & Trending - Each structure or component is inspected from the interior and exterior where accessible. Some structures (or portion of structures) may be inaccessible because of radiological considerations, obstructions or other reasons. Plant specific characteristics, industry experience, and/or testing history of such structures under similar environmental conditions may be evaluated in lieu of actual inspection of the inaccessible areas. Whenever normally inaccessible areas are made accessible (i.e., by excavation or other means) an inspection is performed and the results are documented as part of the Inspection Program for Civil Engineering Structures and Components. Inspections are performed by a team of at least two people. Inspectors are qualified by appropriate training and experience and approved by responsible plant management.

The Inspection Program for Civil Engineering Structures and Components is nominally performed every five years with the exact schedule being established with consideration of refueling outages for each unit. The interval may be increased to a nominal ten-year frequency with appropriate justification based on the structure, environment, and related inspection results.

The inspection will be completed in phases as necessary based on the accessibility of each structure, with the goal of completing the inspection and issuing the report within twelve months of starting the inspection. Structures are monitored in accordance with §50.65 (a)(2) provided there is no significant degradation of the structure. Structures which are determined to be unacceptable are monitored in accordance with the provisions contained in §50.65(a)(1) of the Maintenance Rule.

18.4.1.11 Equipment Staging Building Scope:

transition to NFPA 805 reflect that the Equipment Staging Building is a hardened structure that provides a containment to limit radioactive releases resulting from fire fighting activities. The features of the building credited in this regard are the hardened walls and the sump in the building that limit the spread of the fire and contain the products of fire fighting.

18.4 - 32 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 Screening:

For the purpose of screening, the Equipment Staging Building is included in the categorization of Other Structures, as addressed in the Section 3.5-2 of the License Renewal Application.

Since the only intended functions of this building relate to containment of radiological releases from fire fighting activities, only those structural elements that participate in this function are in the scope of license renewal. Specifically, this pertains to the floor, walls, roof, sump, and exterior structural elements. The Equipment Staging Building also houses ventilation system equipment that is credited with monitoring releases from the building, and isolating if necessary to limit releases. The intended functions ascribed to the Equipment Staging Building are to provide a pressure boundary and/or fission product barrier and structural support and/or shelter to components relied on during certain postulated fire, anticipated transients without scram, and/or station blackout events.

Aging Management Review:

Aging management review of the Equipment Staging Building was accomplished consistent with the methodology in the License Renewal Application, as modified by the revised AMR Table 3.5-2 provided in the letter dated October 28, 2002 (ADAMS Accession No. ML023090324). The Equipment Staging Building is constructed of reinforced concrete. Steel supports and support components are also credited with support of Equipment Staging Building Ventilation System components. Consistent with the methodology in the License Renewal Application, the internal environment is considered to be sheltered, and the external environment is considered to be External (exposed to weather). Boric acid leakage is not applicable. Portions of the foundation are below grade.

Based on the review of Equipment Staging Building design details and consistent with License Renewal Application AMR Table 3.5-2 as modified by letter dated October 28, 2002, applicable aging effects for the Equipment Staging Building are loss of material and cracking. The Inspection Program for Civil Engineering Structures and Components is credited with managing these aging effects. Aging management review for equipment and component supports within the scope of license renewal for the Equipment Staging Building are bounded by the AMR line items identified in License Renewal Application AMR Table 3.5-3, and also credit the Inspection Program for Civil Engineering Structures and Components for aging management of loss of material.

Aging Management Programs:

The Inspection Program For Civil Engineering Structures and Components was an existing program in the License Renewal Application (Section B.3.21). UFSAR Section 18.2.17 contains the following description of the Inspection Program For Civil Engineering Structures and Components:

The Inspection Program for Civil Engineering Structures and Components is intended to meet the requirements of 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants (the Maintenance Rule). This program:

(1) monitors and assesses mechanical components, civil structures and components and their condition in order to provide reasonable assurance that they are capable of performing their intended functions in accordance with the current licensing basis; (2) includes nuclear safety-related structures which enclose, support, or protect nuclear safety-related systems and components, non-safety-related structures whose failure may prevent a nuclear safety-related system or component from fulfilling its intended (13 APR 2020) 18.4 - 33

UFSAR Chapter 18 McGuire Nuclear Station function, and non safety-related structures which support equipment relied on during certain regulated events.

NEI 96-03, Industry Guideline for Monitoring the Condition of Structures at Nuclear Power Plants, has been used as guidance in the preparation of the Inspection Program for Civil Engineering Structures and Components. Examination and assessment of the condition of a structure is performed using guidance provided in codes and standards such as:

  • ACI 349.3, Evaluation of Existing Nuclear Safety-Related Concrete Structures Specific corrective actions are implemented in accordance with the Problem Investigation Process. The Problem Investigation Process applies to all structures and components within the scope of the Inspection Program for Civil Engineering Structures and Components.

Specification DPS-1274.00-00-0005, License Renewal Aging Management Programs and Activities, provides the basis for aging management activities, updated through the receipt of the License Renewal SER. This specification contains a 10 element description of aging management programs, including the Inspection Program For Civil Engineering Structures and Components. Although the UFSAR description of the Inspection Program for Civil Engineering Structures and Components already references RG 1.127, a review of the 10 element program description DPS-1274.00-00-0005 finds the following changes (in bold) are needed to address inspection of the Equipment Staging Building Program

Description:

The Inspection Program for Civil Engineering Structures and Components is credited with managing the following aging effects for the period of extended operation:

  • Loss of material due to corrosion for exposed surfaces of steel components:

anchorage / embedments; cable tray and conduit supports; checkered plates; equipment component supports; expansion anchors; flood, pressure, and specialty doors; HVAC duct supports; instrument line supports; instrument racks and frames; lead shielding supports; metal siding; pipe supports; stair, platform, and grating supports; structural steel beams, columns, plates and trusses; sump screens, Shield Building penetrations; and the unit vent stack

  • Cracking of masonry block walls
  • Change in material properties due to leaching of concrete walls and roofs
  • Loss of material and cracking for reinforced concrete beams, columns, and walls for the Nuclear Service Water and Standby Nuclear Service Water Pump Structure and the Low Pressure Service Water Intake Structure (Catawba only)
  • Cracking and change in material properties of elastomeric flood seals (Catawba only)
  • Loss of material of composite roofing
  • Loss of material of exposed external surfaces of mechanical components (program will be enhanced to add this)
  • Loss of material of the steel components of the Yard Drainage System (program will be enhanced to add this for Catawba only) 18.4 - 34 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18

  • Although Duke did not identify any aging effects for other concrete components beyond those identified above, the NRC staff in its SER dated August 14, 2002 identified loss of material, cracking and change in material property to be both plausible and applicable aging effects for all concrete components. Notwithstanding the disagreement on the aging effects that require management for the period of extended operation, Duke committed, in its response to Open Items 3.5-1 and 3.5-3 provided in letters dated October 2, 2002 and November 14, 2002, to perform periodic inspections of the remaining concrete components to manage the aging effects of loss of material, cracking, and change in material properties using the Inspection Program for Civil Engineering Structures and Components.

The Inspection Program for Civil Engineering Structures and Components is applicable in meeting the regulatory requirements of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants. The Inspection Program for Civil Engineering Structures and Components is a condition monitoring program.

Additional Information: The Maintenance Rule requires that the performance condition of structures be monitored in a manner sufficient to provide reasonable assurance that those structures are capable of fulfilling their intended function. Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, NUMARC 93-01 [Reference 1],

paragraph 9.4.2.4 provides examples of structural monitoring activities including visual inspection and non-destructive examination. NUREG-1526 [Reference 2] states that certain structures such as the primary containment can be monitored through the performance of established testing requirements however, other structures such as reactor buildings, auxiliary buildings, and cooling towers, may be more amenable to condition monitoring. Structures subject to this examination are listed in EDM-410, Inspection Program for Civil Engineering Structures and Components, Table 410-2 [Reference 3]. The inspection program contains more than the structures within license renewal scope because it addresses structures within the scope of the Maintenance Rule.

The Inspection Program for Civil Engineering Structures and Components is evaluated for each program element in accordance with the guidance provided in Appendix C of specification DPS-1274.00-00-0001, License Renewal Technical Information Development Guide [Reference 4] to provide reasonable assurance that the effects of aging will be adequately managed for the components so that their intended function(s) will be maintained consistent with the CLB for the period of extended operation. The evaluation results are provided below.

Scope - The scope of the Inspection Program for Civil Engineering Structures and Components includes the following structures and the exposed external surfaces of mechanical components located within them:

McGuire Nuclear Station

  • Auxiliary Building Structures (including the Control Building, Diesel Generator Buildings, Fuel Buildings, Main Steam Doghouses)
  • Reactor Buildings (including Unit 1 and 2 internal structures, and Station Vents)
  • Standby Shutdown Facility
  • Condenser Cooling Water (RC) Intake Structure (fire pump rooms only)

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UFSAR Chapter 18 McGuire Nuclear Station

  • Turbine Building (including Service Building)
  • Yard Structures (including Refueling Water Storage Tank and Reactor Make-up Water Storage Tank foundations, Refueling Water Storage Tank missile wall, Transformer Station and Switchyard electrical enclosures and equipment pads, Relay House, trenches (pipe and cable), equipment and component supports located outside)
  • Equipment Staging Building Monitoring & Trending - Each structure or component is inspected from the interior and exterior where accessible. Some structures (or portion of structures) may be inaccessible because of radiological considerations, obstructions or other reasons. Plant specific characteristics, industry experience, and/or testing history of such structures under similar environmental conditions may be evaluated in lieu of actual inspection of the inaccessible areas. Whenever normally inaccessible areas are made accessible (i.e., by excavation or other means) an inspection is performed and the results are documented as part of the Inspection Program for Civil Engineering Structures and Components. Inspections are performed by a team of at least two people. Inspectors are qualified by appropriate training and experience and approved by responsible plant management.

The Inspection Program for Civil Engineering Structures and Components is nominally performed every five years with the exact schedule being established with consideration of refueling outages for each unit. The interval may be increased to a nominal ten-year frequency with appropriate justification based on the structure, environment, and related inspection results.

The inspection will be completed in phases as necessary based on the accessibility of each structure, with the goal of completing the inspection and issuing the report within twelve months of starting the inspection. Structures are monitored in accordance with §50.65 (a)(2) provided there is no significant degradation of the structure. Structures which are determined to be unacceptable are monitored in accordance with the provisions contained in §50.65(a)(1) of the Maintenance Rule.

18.4.1.12 Equipment Staging Building Ventilation System - Mechanical Elements Scoping:

The intended function of the Equipment Staging Building Ventilation System is to provide a filtered, monitored release path for smoke and combustion byproducts, in support of radiological release evaluations required for compliance with NFPA 805. As such, only the portions of the system on the release path that participate in this function are included in the scope of license renewal. Those portions of the system that only support personnel / occupancy are not included in scope, and neither are those portions of the system that provide ventilation in areas not housing radioactive materials. Based on a review of system design information, the portion of the system that is included in the scope of license renewal consists of system ducting and ventilation components in the flow path that collect air from various areas of the building and convey it through the exhaust area filter units, to radiation monitoring instrumentation, to the exhaust fan, and then to an elevated release point above the Equipment Staging Building roof.

Screening:

Those mechanical components on the flow path in the scope of license renewal are considered to perform a passive, pressure boundary function and are subject to aging management review.

Those portions of system components in air handlers, dampers, etc., that move or change state to perform the intended function are considered active, and exempt from aging management review. Filter elements are considered short-lived by design, as these components are intended 18.4 - 36 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 to be routinely monitored and serviced as necessary. Those passive, long-lived components in the scope of license renewal that are subject to aging management review are identified in the aging management review table below. The components have a pressure boundary intended function.

Aging Management Review:

Ducting / duct components in the exhaust ducting in the system are constructed of galvanized steel. Flexible connectors are installed on both sides of the inline exhaust fan, and are constructed out of neoprene. The HEPA and charcoal filters are housed in a steel / galvanized steel housing. The internal environment for the duct system through the exhaust point is ventilation system air. The external environment is sheltered.

Aging management reviews of the mechanical portion of the Equipment Staging Building Ventilation System were performed consistent with the methodology of the License Renewal Application. The results of these aging management reviews are presented in the following table:

Component Component Material Internal Aging Effects Aging Type Function Environment Management Program and Activities External Environment Air Handling PB CS/GS Ventilation None None Required Unit & Filter housings (housing only)

Sheltered Ductwork PB GS Ventilation None None Required Sheltered Flexible PB Neoprene Ventilation Connectors None None Required Sheltered Aging Management Programs:

None Required (13 APR 2020) 18.4 - 37

UFSAR Chapter 18 McGuire Nuclear Station 18.4.1.13 Turbine Buildings and Service Building Fire Rated Assemblies Scoping:

Calculation MCC -1435.03-00-0014 Attachment 6 identifies that, under NFPA 805 requirements, the Turbine Buildings and the Service Building became individual fire areas, required to maintain separation with fire barriers with a resistance rated for at least 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

Based on the conclusions of this calculation, the fire barriers separating the Turbine Building, the Service Building, and adjacent structures are required for compliance with NFPA 805 fire protection requirements under 10 CFR 50.48(c). These fire barriers were added to the listing of required fire rated assemblies in SLC 16.9.5 in the MNS transition to an NFPA 805, and are considered to meet the scoping criteria of 10 CFR 54.4(a)(3). Accordingly, these fire barrier features are ascribed a "provides rated fire barrier" intended function.

A review of the MNS License Renewal Application finds that both the Service Building and the Turbine Buildings were initially included in the scope of license renewal. A review of structural commodities subject to aging management review shows that these buildings were addressed under AMR table 3.5-2, Aging Management Review Results - Other Structures. The LRA AMR line items for fire barrier penetrations and fire doors in the Table 3.5-2 limited scoping of these features to just the Auxiliary Building; fire barrier penetrations and fire doors in the Turbine Buildings and Service Building were excluded from scope.

Based on this review, the fire barrier penetrations and fire doors in the Turbine Buildings and Service Building were not initially included in the scope of license renewal, are now assigned intended functions in the NFPA 805 Fire Protection Program, and so are considered newly identified SSCs.

Screening:

The fire barrier penetrations and fire barrier door assemblies in the Turbine Buildings and Service Building are long-lived SSCs considered to perform a passive intended function in providing a rated / approved barrier to the spread of fire.

Aging Management Review:

The fire barrier penetrations and fire doors in the Turbine Buildings and Service Building are comparable in function, design, and service environment to those in the Auxiliary Building. The credited Turbine Buildings and Service Building fire barrier penetrations and fire door assemblies were added to the listing of required fire rated assemblies in SLC 16.9.5 in the MNS transition to an NFPA 805. Therefore, aging management review of the fire barrier penetrations and fire doors in the Turbine Buildings and Service Building will be consistent with their treatment in the Auxiliary Building in LRA Table 3.5-2. The annotation that these AMR lines is applicable only to the Auxiliary Building is revised to reflect applicability to the Turbine Buildings and Service Building, as follows:

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McGuire Nuclear Station UFSAR Chapter 18 Component Component Material Environment Aging Effects Aging Management Type Function Program and Activities Steel Structural Components 4 Steel Sheltered Fire Protection Fire Doors Program External (AB, TB, and SVB) (Yard Only)

Other Structural Components Fire Barrier 4 Silicone Sheltered Fire Protection Penetratio Cracking Program n Seals Separation (AB, TB, and SVB)

Rubber Sheltered Cracking 4 - Provides rated fire barrier to confine or retard fire from spreading to or from adjacent areas of the plant.

Aging Management Programs The MNS Fire Protection Program described in LRA Section B.3.12 consists of two parts: (1)

Fire Barrier Inspections, and (2) Mechanical Fire Protection Components Tests and Inspections.

Of these, Fire Barrier Inspections described in B.3.12.1 is applicable to fire doors and fire barrier penetrations, and is credited with managing aging effects of these SSCs in the Turbine Buildings and Service Building. The Fire Barrier Inspections are driven by requirements in SLC 16.9.5, Fire Rated Assemblies, and described in the UFSAR by reference to the applicable SLC in Table 18-1. The fire doors and fire barrier penetrations in the Turbine Buildings and Service Building that are credited in the transition to an NFPA 805 Fire Protection Program have been added to this SLC, as applicable. Therefore, no further updates are needed to the UFSAR description of the Fire Protection Program - Fire Barrier Inspections, to address aging management of the fire doors and fire barrier penetrations in the Turbine Buildings and Service Building.

18.4.1.14 Control Rod Drive Mechanism Penetration Thermal Sleeves Control Rod Drive Mechanism Penetration Thermal Sleeves - Mechanical Elements Scoping:

A Westinghouse 10 CFR Part 21 Notification dated May 23, 2018 identified that wear of the Control Rod Drive Mechanism (CRDM) Penetration Thermal Sleeves presents a safety concern regarding the potential to interfere with control rod movement. Based on the applicability of this (13 APR 2020) 18.4 - 39

UFSAR Chapter 18 McGuire Nuclear Station notification, the CRDM Penetration Thermal Sleeves are considered to meet the scoping criteria of 10 CFR 54.4(a)(2), and are considered newly identified SSCs.

Screening:

The CRDM Penetration Thermal Sleeves are passive components, which must maintain structural / dimensional integrity to preclude interference with the movement of the control rods.

Aging Management Review:

The aging effect of loss of material due to wear of the CRDM Penetration Thermal Sleeves will be managed by augmented inspections under the lnservice Inspection Program.

Component Component Material Internal Aging Aging Type Function Environment Effects Management Program and Activities External Environment Control Rod Structural Stainless Treated Loss of Inservice Drive Integrity Steel / Water Material Inspection Penetration Nickel Alloy Due to Wear Program Thermal Sleeves Treated Water Aging Management Programs:

Subsequent to the 10 CFR Part 21 notification pertaining to CRDM Penetration Thermal Sleeve wear, Westinghouse issued Nuclear Safety Advisory Letter NSAL-18-1, which provided affected licensees a basis for operation and inspection recommendations. As discussed in the NSAL, portions of previous vessel internals inspection guidance in TB-07-2 Revision 3 were superseded, and Westinghouse had worked with the Pressurized Water Reactor Owners Group (PWROG) to develop acceptance criteria (contained in PWROG-16003-P) to be used by the participating members when measuring thermal sleeve flange wear. Duke's review of NSAL-18-1 resulted in follow up actions, including performance of thermal sleeve flange wear measurements using the criteria of PWROG-16003-P. Notably, MNS Unit 2 inspections performed during M2R25 (Fall 2018) identified that flange separation had occurred at one location (H8), and a thermal sleeve replacement was performed. The remainder of the thermal sleeves did not exceed the separation criteria provided in PWROG- 16003-P. Unit 1 thermal sleeve inspections were performed during 1R26 (Spring 2019), with no adverse findings identified.

Consistent with the recommendations of the NSAL, programmatic activities for thermal sleeve inspections have been incorporated into the Inservice Inspection Program as augmented inspections. Accordingly, wear of the CRDM thermal sleeves will be managed by ongoing augmented inspections under the Inservice Inspection Program. The Inservice Inspection Program was an existing program in the MNS LRA (Ref. Section B.3.21), and is described in 18.4 - 40 (13 APR 2020)

McGuire Nuclear Station UFSAR Chapter 18 UFSAR Section 18.2.16. The following additions (in bold) are made to the Inservice Inspection Program UFSAR description to address aging management of the CRDM thermal sleeves:

18.2.16 Inservice Inspection Plan The McGuire Inservice Inspection Plan, implements the requirements of 10 CFR 50.55a for Class 1, 2, and 3 components and Class 1, 2, 3, and MC component supports. The examinations are performed to the extent practicable within the limitations of design, geometry and materials of construction of the component. The period of extended operation for McGuire will contain the 5th and 6th ten-year inservice inspection intervals.

The Inservice Inspection Plan includes the following inspections and activities:

  • McGuire Unit 1 Cold Leg Elbow
  • A VT-1 examination of the reactor vessel internals clevis insert fasteners will be performed in lieu of VT-3 examination currently required by ASME Section XI. (Note

- Duke has submitted a revision to the McGuire Reactor Vessel Internals Inspection Program based on the requirements of MRP-227-A [References 28 and 29]. Once this submittal is approved by the NRC, inspection of reactor vessel internals clevis insert fasteners will be performed in compliance with this updated program, and the UFSAR description of the McGuire Inservice Inspection Plan will be updated accordingly.

18.2.16.3 Control Rod Drive Mechanism Penetration Thermal Sleeve Inspections Wear of the Control Rod Drive Mechanism Penetration Thermal Sleeves is an aging effect that can lead to interference with control rod movement. The Inservice Inspection Program includes augmented inspections of the thermal sleeves to determine wear measurements, and trending to ensure that corrective actions are taken prior to failure of the thermal sleeve flange.

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McGuire Nuclear Station UFSAR Chapter 18 18.5 Newly Identified TLAAs The following TLAAs associated with the 40 - 60 year period of extended operation were identified after receipt of the renewed operating license.

18.5.1 Upflow Modification TLAA Five TLAAs associated with the 40 - 60 year period of extended operation were identified during the development of the Reactor Vessel Internals inspection plan. Three TLAAs were identified related to fatigue usage factors (one each for the modified top former plate (with holes), the core barrel plugs, and the lower former plate plugs) and two TLAAs related to Irradiation-Enhanced Stress Relaxation (one each for the core barrel plugs and the lower former plate plugs.)

Regarding fatigue usage for the modified top former plate (with holes), the core barrel plugs, and the lower former plate plugs, it was concluded that the transient design cycles for 60 years are assumed to be equal to the design cycles for 40 years, and a change in fatigue of the reactor vessel internals need not be addressed. Accordingly, these TLAAs are resolved under the criterion of 10 CFR 54.21(c)(1)(i).

TLAAs related to Irradiation-Enhanced Stress Relaxation (ISR) were resolved by consideration of the reduction of end-of-life (EOL) interface pressure between the plug and its respective item (i.e., core barrel or lower former plate) from the beginning-of-life (BOL) results due to short-term and long-term thermal stress relaxation (SR) and ISR. The minimum required stress ratio for the core barrel and lower former plate plugs was calculated for 60 years. This minimum required stress ratio was based on loads and calculations of EOL interface pressure (core barrel plugs),

minimum required test plug spring back pressure (lower former plate plugs), and required contact pressure (core barrel and lower former plate plugs). The minimum required stress ratio values were compared to and determined to be less than the respective estimated stress ratio for both the core barrel and lower former plate plugs. Therefore, the core barrel and lower former plate plugs meet the requirements to remain in place for 60 years of plant life.

Accordingly, these TLAAs are resolved under the criterion of 10 CFR 54.21(c)(1)(ii).

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