ML18247A272

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Revision 36 to Final Safety Analysis Report, Chapter 19, License Renewal
ML18247A272
Person / Time
Site: Millstone  Dominion icon.png
Issue date: 06/18/2018
From:
Dominion Energy Nuclear Connecticut
To:
Office of Nuclear Reactor Regulation
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ML18199A125 List:
References
18-225
Download: ML18247A272 (51)
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Text

Millstone Power Station Unit 3 Safety Analysis Report Chapter 19: License Renewal

Table of Contents tion Title Page INTRODUCTION .................................................................................... 19.0-1 AGING MANAGEMENT........................................................................ 19.1-1

.1 Aging Management Programs .................................................................. 19.1-1

.2 Time Limited Aging Analyses Aging Management Programs: ............... 19.1-2

.3 References for Section 19.1 ...................................................................... 19.1-2 PROGRAMS THAT MANAGE THE EFFECTS OF AGING ON STRUCTURES AND COMPONENTS WITHIN THE SCOPE OF LICENSE RENEWAL .......................................................... 19.2-1

.1 Aging Management Programs .................................................................. 19.2-1

.1.1 Battery Rack Inspections .......................................................................... 19.2-1

.1.2 Boric Acid Corrosion................................................................................ 19.2-2

.1.3 Buried Pipe Inspection Program ............................................................... 19.2-2 2.1.4 Chemistry Control for Primary Systems Program .................................... 19.2-3

.1.5 Chemistry Control for Secondary Systems Program ................................ 19.2-4

.1.6 Closed-Cycle Cooling Water System ....................................................... 19.2-4

.1.7 Electrical Cables and Connectors Not Subject to 10 CFR 50.49 Environmental Qualification Requirements ...................................................................... 19.2-5

.1.8 Electrical Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits ....................................... 19.2-6

.1.9 Fire Protection Program............................................................................ 19.2-7

.1.10 Flow-Accelerated Corrosion..................................................................... 19.2-8

.1.11 Fuel Oil Chemistry.................................................................................... 19.2-8

.1.12 General Condition Monitoring.................................................................. 19.2-9

.1.13 Inaccessible Medium Voltage Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements ........................................... 19.2-10

.1.14 Infrequently Accessed Areas Inspection Program.................................. 19.2-11

.1.15 Inservice Inspection Program: Containment Inspections ....................... 19.2-12

.1.16 Inservice Inspection Program: Reactor Vessel Internals ........................ 19.2-12

.1.17 Inservice Inspection Program: Systems, Components and Supports...... 19.2-14 28/18 19-i Rev. 31

tion Title Page

.1.18 Inspection Activities: Load Handling Cranes and Devices .................... 19.2-15 2.1.19 Reactor Vessel Surveillance ................................................................... 19.2-16

.1.20 Service Water System (Open-Cycle Cooling) ........................................ 19.2-16

.1.21 Steam Generator Structural Integrity ...................................................... 19.2-17

.1.22 Structures Monitoring Program .............................................................. 19.2-18

.1.23 Tank Inspection Program........................................................................ 19.2-19

.1.24 Work Control Process ............................................................................. 19.2-20

.1.25 Bolting Integrity Program ....................................................................... 19.2-22

.2 References for Section 19.2 .................................................................... 19.2-22 TIME-LIMITED AGING ANALYSIS .................................................... 19.3-1 3.1 Reactor Vessel Neutron Embrittlement .................................................... 19.3-1

.1.1 Upper Shelf Energy .................................................................................. 19.3-1

.1.2 Pressurized Thermal Shock ...................................................................... 19.3-1

.1.3 Pressure-Temperature Limits.................................................................... 19.3-2

.2 Metal Fatigue ............................................................................................ 19.3-2

.2.1 Millstone Unit 3 Class 1 Components ...................................................... 19.3-3

.2.2 Non-Class 1 Components ......................................................................... 19.3-4

.2.3 Environmentally Assisted Fatigue ............................................................ 19.3-4

.3 Environmental Qualification (EQ) of Electric Equipment ....................... 19.3-6

.4 Containment Liner Plate, Metal Containments, and Penetrations Fatigue Anal-ysis ............................................................................................................ 19.3-6

.4.1 Containment Liner Plate ........................................................................... 19.3-6 3.4.2 Containment Penetrations ......................................................................... 19.3-6

.5 Other Plant-Specific Time-Limited Aging Analyses................................ 19.3-7

.5.1 Crane Load Cycle Limit ........................................................................... 19.3-7

.5.2 Reactor Coolant Pump Flywheel .............................................................. 19.3-7

.5.3 Leak-Before-Break ................................................................................... 19.3-8

.5.4 Containment Subfoundation ..................................................................... 19.3-9

.6 References for Section 19.3 ...................................................................... 19.3-9 28/18 19-ii Rev. 31

tion Title Page TLAA SUPPORT PROGRAMS .............................................................. 19.4-1

.1 Electrical Equipment Qualification .......................................................... 19.4-1

.2 Metal Fatigue of Reactor Coolant Pressure Boundary ............................. 19.4-1 EXEMPTIONS ......................................................................................... 19.5-1 LICENSE RENEWAL COMMITMENTS .............................................. 19.6-1

.1 References for Section 19.6 ...................................................................... 19.6-1 28/18 19-iii Rev. 31

List of Tables mber Title

-1 License Renewal Commitments 28/18 19-iii Rev. 31

INTRODUCTION application for a renewed operating license is required by 10 CFR 54.21(d) to include a R Supplement. This appendix, which includes the following sections, comprises the FSAR plement:

Section 19.1 contains a listing of the aging management programs and the status of the program at the time the License Renewal Application was submitted.

Section 19.2 contains a description of the programs for managing the effects of aging.

Section 19.3 contains the evaluation of Time-limited Aging Analyses (TLAAs) for the period of extended operation.

Section 19.4 contains a summarized description of the programs that support the TLAAs.

Section 19.5 contains a summarized description of the plant-specific exemptions.

Section 19.6 contains a matrix of the license renewal commitments.

integrated plant assessment for license renewal identified new and existing aging agement programs necessary to provide reasonable assurance that components within the pe of license renewal will continue to perform their intended functions consistent with the rent Licensing Basis (CLB) for the period of extended operation. The period of extended ration is defined as 20 years from the units previous 40 year operating license expiration date.

ess otherwise identified, references to the Operating License are considered a reference to the ewed Operating License.

28/18 19.0-1 Rev. 31

.1 AGING MANAGEMENT PROGRAMS aging management programs for Millstone Unit3 are described in the following sections. The grams are either consistent with generally accepted industry methods as discussed in REG-1801 (Reference 19.1-1), require enhancements to be consistent with generally accepted ustry standards, or are site-specific programs.

following list reflects the status of these programs at the time this section was included in the R and provides a historical perspective of their status at the completion of the NRC review of License Renewal Application. The implementation status of the listed programs will change ew programs are developed and enhancements to existing programs are completed.

mmitments for program additions and enhancements are identified in the appropriate sections.

1. Battery Rack Inspections (Section 19.2.1.1) (Existing - Requires Enhancement).
2. Boric Acid Corrosion (Section 19.2.1.2) (Existing).
3. Buried Pipe Inspection Program (Section 19.2.1.3) (Existing - Requires Enhancement).
4. Chemistry Control for Primary Systems Program (Section 19.2.1.4) (Existing).
5. Chemistry Control for Secondary Systems Program (Section 19.2.1.5) (Existing).
6. Closed-Cycle Cooling Water System (Section 19.2.1.6) (Existing - Requires Enhancement.)
7. Electrical Cables and Connectors Not Subject to 10 CFR 50.49 Environmental Qualification Requirements (Section 19.2.1.7) (To Be Developed).
8. Electrical Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits (Section 19.2.1.8)

(Existing - Requires Enhancement).

9. Fire Protection Program (Section 19.2.1.9) (Existing - Requires Enhancement).
10. Flow-Accelerated Corrosion (Section 19.2.1.10) (Existing).
11. Fuel Oil Chemistry (Section 19.2.1.11) (Existing).
12. General Condition Monitoring (Section 19.2.1.12) (Existing - Requires Enhancement).

28/18 19.1-1 Rev. 31

Enhancement).

14. Infrequently Accessed Areas Inspection Program (Section 19.2.1.14) (To Be Developed).
15. Inservice Inspection Program: Containment Inspections (Section 19.2.1.15)

(Existing).

16. Inservice Inspection Program: Reactor Vessel Internals (Section 19.2.1.16)

Existing - Requires Enhancement).

17. Inservice Inspection Program: Systems, Components and Supports (Section 19.2.1.17) (Existing - Requires Enhancement).
18. Inspection Activities: Load Handling Cranes and Devices (Section 19.2.1.18)

(Existing - Requires Enhancement).

19. Reactor Vessel Surveillance (Section 19.2.1.19) (Existing).
20. Service Water System (Open-Cycle Cooling) (Section 19.2.1.20) (Existing).
21. Steam Generator Structural Integrity (Section 19.2.1.21) (Existing).
22. Structures Monitoring Program (Section 19.2.1.22) (Existing - Requires Enhancement).
23. Tank Inspection Program (Section 19.2.1.23) (Existing - Requires Enhancement).
24. Work Control Process (Section 19.2.1.24) (Existing - Requires Enhancement).
25. Bolting Integrity Program (Section 19.2.1.25) (Existing)

.2 TIME LIMITED AGING ANALYSES AGING MANAGEMENT PROGRAMS:

1. Electrical Equipment Qualification (Section 19.4.1) (Existing).
2. Metal Fatigue of Reactor Coolant Pressure Boundary (Section 19.4.2) (Existing).

.3 REFERENCES FOR SECTION 19.1

-1 NUREG-1801, Generic Aging Lessons Learned (GALL) Report, U. S. Nuclear Regulatory Commission, July 2001.

28/18 19.1-2 Rev. 31

s section provides summaries of the programs credited for managing the effects of aging on ctures and components within the scope of license renewal.

Quality Assurance Program implements the requirements of 10 CFR 50, Appendix B, and is sistent with the summary in NUREG-1800, Section A.2. The Quality Assurance program udes the elements of corrective action, confirmation process, and administrative controls and pplicable to the safety-related and non-safety-related structures, and components that are hin the scope of license renewal.

.1 AGING MANAGEMENT PROGRAMS

.1.1 Battery Rack Inspections gram Description tery Rack Inspections is a plant-specific program that manages the aging effect of loss of erial. The structural integrity of the support racks for the station batteries, within the scope of nse renewal, is verified by visually inspecting for loss of material.

acceptance criterion for visual inspections is the absence of anomalous indications that are s of degradation. Corrective actions for conditions that are adverse to quality are performed in ordance with the Corrective Action Program as part of the Quality Assurance Program. The ective action process provides reasonable assurance that deficiencies adverse to quality are er promptly corrected or are evaluated to be acceptable.

mmitments following program enhancements will be implemented prior to the period of extended ration:

Inclusion of In-Scope Battery Racks The existing inspection program will be modified to include those battery racks that require monitoring for license renewal, but are not already included in the program. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 1.

Inspection Criteria Implementing procedures will be modified to include loss of material as a potential aging effect and to provide guidance on the inspection of items (such as anchorages, bracing and supports, side and end rails, and spacers), which contribute to battery rack integrity or seismic design of the battery racks. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 2.

28/18 19.2-1 Rev. 31

gram Description ic Acid Corrosion corresponds to NUREG-1801,Section XI.M10 Boric Acid Corrosion.

program manages the aging effect of loss of material and ensures that systems, structures, and ponents susceptible to boric acid corrosion are properly monitored. The program uses visual ections to detect the boric acid leakage source, path, and any targets of the leakage. It ensures boric acid corrosion is consistently identified, documented, evaluated, trended, and ctively repaired. The Boric Acid Corrosion program provides both detection and analysis of age of borated water inside containment. The General Condition Monitoring program is the ary method for detecting borated water leakage outside containment. The analysis of the age is performed through the Boric Acid Corrosion program. Any necessary corrective ons are implemented through the Corrective Action Program.

ic Acid Corrosion program implements the requirements of:

NRC Bulletin 2001-01 (Reference 19.2-15).

NRC Bulletin 2002-01 (Reference 19.2-16).

NRC Bulletin 2002-02 (Reference 19.2-17).

NRC Bulletin 2003-02 (Reference 19.2-18).

NRC Order EA-03-009 (Reference 19.2-19).

NRC Bulletin 2004-01 (Reference 19.2-20).

acceptance criterion is the absence of any boric acid leakage or precipitation. If boric acid age or precipitation is found by any personnel, it is required to be reported using the rective Action Program. Corrective actions for conditions that are adverse to quality are ormed in accordance with the Corrective Action Program as part of the Quality Assurance gram. The corrective action process provides reasonable assurance that deficiencies adverse to lity are either promptly corrected or are evaluated to be acceptable.

.1.3 Buried Pipe Inspection Program gram Description Buried Pipe Inspection Program is an existing program that corresponds to NUREG-1801, tions XI.M28, Buried Piping and Tanks Surveillance and XI.M34, Buried Piping and ks Inspection. The program manages the aging effect of loss of material through the use of ventive measures and inspections. The inspections will be performed when the piping and ponents are excavated for maintenance or for any other reason.

28/18 19.2-2 Rev. 31

acceptance criterion for visual inspections is the absence of anomalous indications that are s of degradation. In addition to visual inspections, the field inspections for loss of material to selective leaching will include mechanical means, such as resonance when struck by ther object, scraping, or chipping. Corrective actions for conditions that are adverse to quality performed in accordance with the Corrective Action Program as part of the Quality Assurance gram. The corrective action process provides reasonable assurance that deficiencies adverse to lity are either promptly corrected or are evaluated to be acceptable.

mmitments following program enhancements will be implemented prior to the period of extended ration:

Baseline Inspection A baseline inspection of the in-scope buried piping located in a damp soil environment will be performed for a representative sample of each combination of material and protective measures. Inspection for the loss of material due to selective leaching will be performed by visual, and mechanical or other appropriate methods. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 3.

Buried Piping Inspections The maintenance and work control procedures will be revised to ensure that inspections of buried piping are performed when the piping is excavated during maintenance or for any other reason. These procedures will include the inspection for the loss of material due to selective leaching, which will be performed by visual, and mechanical or other appropriate methods. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 4.

.1.4 Chemistry Control for Primary Systems Program gram Description mistry Control for Primary Systems Program corresponds to NUREG-1801,Section XI.M2, ater Chemistry. The program includes periodic monitoring and control of known detrimental taminants such as chlorides, fluorides, dissolved oxygen, and sulfate concentrations below the ls known to result in loss of material or cracking. Water chemistry control is in accordance h the guidelines in EPRI TR-105714 (Reference 19.2-1) for primary water chemistry.

acceptance criterion is that the maximum levels for the monitored contaminants are ntained below the system-specific limits. Corrective actions for conditions that are adverse to lity are performed in accordance with the Corrective Action Program as part of the Quality 28/18 19.2-3 Rev. 31

.1.5 Chemistry Control for Secondary Systems Program gram Description mistry Control for Secondary Systems Program corresponds to NUREG-1801, Section M2, Water Chemistry. The program includes periodic monitoring and control of known imental contaminants such as chlorides, sodium, dissolved oxygen, and sulfate concentrations w the levels known to result in loss of material or cracking. Water chemistry control is in ordance with the guidelines in EPRI TR-102134 (Reference 19.2-2) for secondary water mistry.

acceptance criterion is that the maximum levels for the monitored contaminants are ntained below the system-specific limits. Corrective actions for conditions that are adverse to lity are performed in accordance with the Corrective Action Program as part of the Quality urance Program. The corrective action process provides reasonable assurance that deficiencies erse to quality are either promptly corrected or are evaluated to be acceptable.

.1.6 Closed-Cycle Cooling Water System gram Description sed-Cycle Cooling Water System corresponds to NUREG-1801,Section XI. M21, Closed-le Cooling Water System. The program manages the aging effect of loss of material through maintenance of process fluid chemistry and performance monitoring of closed-cycle cooling er systems to ensure parameters remain within acceptable limits. The program is based ctly on guidance contained in EPRI Report TR-107396 (Reference 19.2-3).

acceptance criterion is that the maximum levels for the monitored contaminants are ntained below the system specific limits. Corrective actions for conditions that are adverse to lity are performed in accordance with the Corrective Action Program as part of the Quality urance Program. The corrective action process provides reasonable assurance that deficiencies erse to quality are either promptly corrected or are evaluated to be acceptable.

mmitments following commitment will be implemented prior to the period of extended operation:

Heat Exchanger Baseline Inspection A baseline visual inspection will be performed of the accessible areas of the shell side (including accessible portions of the exterior side of the tubes) of one:

  • Millstone Unit 2 Reactor Building Closed Cooling Water heat exchanger, 28/18 19.2-4 Rev. 31

s commitment is identified in Table 19.6-1, License Renewal Commitments, Item 30.

.1.7 Electrical Cables and Connectors Not Subject to 10 CFR 50.49 Environmental Qualification Requirements gram Description ctrical Cables and Connectors Not Subject to 10 CFR 50.49 Environmental Qualification uirements corresponds to NUREG-1801,Section XI.E1, Electrical Cables and Connections Subject to 10 CFR 50.49 Environmental Qualification Requirements as modified by NRC rim Staff Guidance-05 (Reference 19.2-4). This program manages the aging effects of king and embrittlement to ensure that electrical cables, connectors, and fuse holders within scope of license renewal that are exposed to an adverse localized environment (but not subject he environmental qualification requirements of 10 CFR 50.49) are capable of performing their nded function. Adverse localized environments may be caused by heat, radiation or moisture.

acceptance criterion for the visual inspections of accessible non-EQ cable jackets and nector coverings is the absence of anomalous indications that are signs of degradation.

rective actions for conditions that are adverse to quality are performed in accordance with the rective Action Program as part of the Quality Assurance Program. The corrective action cess provides reasonable assurance that deficiencies adverse to quality are either promptly ected or are evaluated to be acceptable.

mmitments following actions will be implemented prior to the period of extended operation:

Program Implementation The Electrical Cables and Connectors Not Subject to 10 CFR 50.49 Environmental Qualification Requirements program will be established.

This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 5.

Inclusion of In-Scope Fuse Holders Fuse holders meeting the requirements will be evaluated prior to the period of extended operation for possible aging effects requiring management. The fuse holder will either be replaced, modified to minimize the aging effects, or this program will manage the aging effects. The Electrical Cables and Connectors Not Subject to 10 CFR 50.49 28/18 19.2-5 Rev. 31

This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 6.

.1.8 Electrical Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits gram Description ctrical Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used nstrumentation Circuits corresponds to NUREG-1801,Section XI.E2, Electrical Cables not ject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation uits and the program as modified in draft NRC ISG-15 (Reference 19.2-5). This program ages the aging effects of cracking and embrittlement for electrical cables within the scope of nse renewal that are used in circuits with sensitive, low-level signals, such as radiation nitoring and nuclear instrumentation (but not subject to the environmental qualification uirements of 10 CFR 50.49), and are installed in adverse localized environments caused by t, radiation or moisture.

acceptance criterion for the calibration readings is the loop-specific tolerances established in hnical Specifications and surveillance procedures. Where calibration of the instrumentation is performed in situ, the acceptance criteria for each test are defined by the specific type of test ormed and the specific cable tested. Corrective actions for conditions that are adverse to lity are performed in accordance with the Corrective Action Program as part of the Quality urance Program. The corrective action process provides reasonable assurance that deficiencies erse to quality are either promptly corrected or are evaluated to be acceptable.

mmitments following program enhancements will be implemented prior to the period of extended ration:

Testing of Cables for Instruments That Are Not Calibrated In Situ Procedures will be developed to employ an alternate testing methodology to confirm the condition of cables and connectors in circuits that have sensitive, low level signals and where the instrumentation is not calibrated in situ. The first tests will be completed prior to the period of extended operation. The frequency of subsequent tests will be based on Engineering evaluation and will not exceed a 10 year interval. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 7.

Review of Surveillance Test Results for Cables Tested In Situ Calibration results for cables tested in situ will be reviewed to detect severe aging degradation of the cable insulation. The initial review will be completed prior to the period 28/18 19.2-6 Rev. 31

This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 33.

.1.9 Fire Protection Program gram Description Fire Protection Program is an existing program and corresponds to NUREG-1801, Sections M26, Fire Protection and XI.M27, Fire Water System and to the revised XI.M27, Fire er System program described in NRC Interim Staff Guidance (ISG)-04 (Reference 19.2-6).

program manages the aging effects of loss of material, cracking, and change of material perties for plant fire protection features and components. The program manages these aging cts through the use of periodic inspections and tests.

program also manages the aging effects for the diesel-driven fire pump fuel supply line, the tor coolant pump oil collection systems, and Appendix R support equipment.

ual inspection of fire protection piping internal surfaces that are exposed to water is performed n the system is opened for maintenance and/or repair. The Work Control Process provides dance for the performance of internal inspections of fire protection piping and components never the system is opened for maintenance or repair.

acceptance criteria for the Fire Protection Program are:

For visual inspections, the absence of anomalous indications that are signs of degradation.

For fire barriers and fire doors, the sizes for breaks, holes, cracks, spalling gaps, and/or clearances are in accordance with the limits established in the inspection procedures.

For fire protection equipment performance tests (i.e., flow and pressure tests), acceptance criteria are provided in the appropriate surveillance procedures.

itionally, the fire protection water system pressure is continuously monitored to be above the imum setpoint. Corrective actions for conditions that are adverse to quality are performed in ordance with the Corrective Action Program as part of the Quality Assurance Program. The ective action process provides reasonable assurance that deficiencies adverse to quality are er promptly corrected or are evaluated to be acceptable.

mmitments following program enhancement will be implemented prior to the period of extended ration:

Baseline Fire Protection Inspections 28/18 19.2-7 Rev. 31

to confirm there is no degradation. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 8.

following program enhancement will be implemented prior to the sprinkler heads achieving ears of service life:

Testing or Replacement of Sprinkler Heads Testing a representative sample of fire protection sprinkler heads or replacing those that have been in service for 50 years will be included in the Fire Protection Program. The first tests will be completed prior to the sprinkler heads achieving 50 years of service life. The frequency of subsequent tests will not exceed a 10 year interval. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 9.

.1.10 Flow-Accelerated Corrosion gram Description w-Accelerated Corrosion Program corresponds to NUREG-1801,Section XI.M17, Flow-elerated Corrosion. The program manages the aging effect of loss of material in accordance h the EPRI guidelines in NSAC-202L (Reference 19.2-7). It includes procedures or inistrative controls to assure that the structural integrity of carbon steel and low-alloy steel ng and components, such as valves, steam traps, and feedwater heaters, is maintained.

engineering evaluations determine if a component needs to be repaired/replaced or is eptable for continued operation until the next scheduled inspection. Corrective actions for ditions that are adverse to quality are performed in accordance with the Corrective Action gram as part of the Quality Assurance Program. The corrective action process provides onable assurance that deficiencies adverse to quality are either promptly corrected or are luated to be acceptable.

2.1.11 Fuel Oil Chemistry gram Description l Oil Chemistry corresponds to NUREG-1801,Section XI.M30, Fuel Oil Chemistry. The gram manages the aging effect of loss of material by monitoring and controlling fuel oil lity to ensure that it is compatible with the materials of construction for in-scope components taining diesel fuel oil.

Fuel Oil Chemistry program uses the following industry standards as the basis for the gram:

ASTM Standard D 1796 (Reference 19.2-8),

28/18 19.2-8 Rev. 31

ASTM Standard D 4057 (Reference 19.2-10).

acceptance criterion is adherence to the specific guidelines and limits defined in related plant cedures for parameters that have been shown to contribute to component degradation.

rective actions for conditions that are adverse to quality are performed in accordance with the rective Action Program as part of the Quality Assurance Program. The corrective action cess provides reasonable assurance that deficiencies adverse to quality are either promptly ected or are evaluated to be acceptable.

.1.12 General Condition Monitoring gram Description eral Condition Monitoring is a plant-specific program that manages the aging effects of loss aterial, change of material properties, and cracking on the external surfaces of components. It erformed in accessible plant areas for components and structures including those within the pe of license renewal and involves visual inspections for evidence of age-related degradation.

eral Condition Monitoring is implemented by Radiation Protection technicians, System ineers, and Plant Equipment Operators while performing their routine in-plant activities.

acceptance criterion for visual inspections is the absence of anomalous indications that are s of degradation. Corrective actions for conditions that are adverse to quality are performed in ordance with the Corrective Action Program as part of the Quality Assurance Program. The ective action process provides reasonable assurance that deficiencies adverse to quality are er promptly corrected or are evaluated to be acceptable.

mmitments following program enhancement will be implemented prior to the period of extended ration:

Procedure and Training Enhancements The procedures and training for personnel performing General Condition Monitoring inspections and walkdowns will be enhanced to provide expectations that identify the requirements for the inspection of aging effects. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 10.

28/18 19.2-9 Rev. 31

gram Description cessible Medium Voltage Cables Not Subject to 10 CFR 50.49 Environmental Qualification uirements corresponds to NUREG-1801,Section XI.E3, Inaccessible Medium-Voltage les not Subject to 10 CFR 50.49 Environmental Qualification Requirements. This program ages the aging effect of formation of water trees and ensures that inaccessible medium-age (2 kV to 15 kV) electrical cables within the scope of license renewal (but not subject to environmental qualification requirements of 10 CFR 50.49) that have been submerged, remain able of performing their intended function. The program considers the combined effects of mergence, simultaneous with a significant voltage exposure. Significant voltage exposure is ned as being subjected to system voltage for more the twenty-five percent of the time.

acceptance criterion for the inspections performed under the Structures Monitoring Program confirm that in-scope, medium-voltage cables have not become submerged. In-scope cable nd to be submerged in standing water for an extended period of time will be subject to an ineering evaluation and corrective action. The evaluation will be based on appropriate testing ng available technology consistent with NRC positions) of cables that are determined to be ted for a significant period of time. The test will use a proven methodology for detecting rioration of the insulation due to wetting. Testing will have acceptance criteria defined in ordance with the specific test identified. Occurrence of degradation that is adverse to quality is red into the Corrective Action Program. The corrective action process provides reasonable rance that deficiencies adverse to quality are either promptly corrected or are evaluated to be eptable.

mmitments following program enhancements will be implemented prior to the period of extended ration:

Verification Testing In-scope cable found to be submerged will be subject to an engineering evaluation and corrective action. The evaluation of cables having significant voltage found to be submerged in standing water for an extended period of time will be based on appropriate testing (using available technology consistent with NRC positions) of cables that are determined to be wetted for a significant period of time. The Engineering evaluation will also address the appropriate testing requirements for the corresponding ten-year intervals during the period of extended operation. The test will use a proven methodology for detecting deterioration of the insulation system due to wetting. Examples of such tests include power factor, partial discharge, or polarization index, as described in EPRI TR-103834-P1-2, Effects of Moisture on the Life of Power Plant Cables, or other appropriate testing. Testing will have acceptance criteria defined in accordance with the specific test identified. Occurrence of degradation that is adverse to quality is entered into 28/18 19.2-10 Rev. 31

Testing of Inaccessible Medium Voltage Cables The in-scope cables in Unit 3 duct lines # 929 (SBO Diesel to Unit 3 4.16kV Normal Switchgear) and # 973 (RSST 3RTX-XSR-B to 6.9kV Normal Switchgear Bus 35A, 35B, 35C and 35D) have been tested to demonstrate that water treeing will not prevent the cables from performing their intended function. Subsequent testing has been scheduled to be performed on a frequency not to exceed a 10 year interval.

This completes the actions required to complete commitment Item 34. in Table 19.6-1, License Renewal Commitments.

Sample Testing of Inaccessible Medium Voltage Cables Prior to the period of extended operation, a representative sample of in-scope medium voltage cables will be tested to demonstrate that water treeing will not prevent the cables from performing their intended function. This sample testing is in addition to the testing specified in the previous commitment. Subsequent testing will be performed on a frequency not to exceed a 10-year interval. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 35.

.1.14 Infrequently Accessed Areas Inspection Program gram Description equently Accessed Areas Inspection Program is a plant-specific program that manages the g effects of loss of material, change of material properties, and cracking. The program uses al inspections of the external surfaces of in-scope structures and components located in equently accessed areas of the plant.

acceptance criterion for visual inspections is the absence of anomalous indications that are s of degradation. Corrective actions for conditions that are adverse to quality are performed in ordance with the Corrective Action Program as part of the Quality Assurance Program. The ective action process provides reasonable assurance that deficiencies adverse to quality are er promptly corrected or are evaluated to be acceptable.

mmitments following program enhancements will be implemented prior to the period of extended ration:

Program Implementation The Infrequently Accessed Areas Inspection Program will be established.

28/18 19.2-11 Rev. 31

.1.15 Inservice Inspection Program: Containment Inspections gram Description rvice Inspection Program: Containment Inspections corresponds to the following NUREG-1 program descriptions:

Section XI.S1, ASME Section XI, Subsection IWE,Section XI.S2, ASME Section XI, Subsection IWL, and Section XI.S4, 10 CFR Part 50, Appendix J.

program manages the aging effects of loss of material, change of material properties, and king. The program is consistent with ASME Section XI, Subsections IWE and IWL, and CFR 50.55a(b)(2), which provide the criteria for ISI Containment inspections.

endix J Leakage Rate Testing is included as part of the Inservice Inspection Program:

tainment Inspections. The Containment Appendix J Leakage Rate Test Program implements e A and B tests to measure the overall primary Containment integrated leakage rate.

acceptance criteria for examinations performed in accordance with the Inservice Inspection gram: Containment Inspections are based on the applicable regulations and standards.

rective actions for conditions that are adverse to quality are performed in accordance with the rective Action Program as part of the Quality Assurance Program. The corrective action cess provides reasonable assurance that deficiencies adverse to quality are either promptly ected or are evaluated to be acceptable.

.1.16 Inservice Inspection Program: Reactor Vessel Internals gram Description rvice Inspection Program: Reactor Vessel Internals corresponds to the following NUREG-1 program descriptions:

Section XI.M12, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS).

Section XI.M13, Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS).

Section XI.M16, PWR Vessel Internals.

28/18 19.2-12 Rev. 31

ension and loss of fracture toughness (which presents itself as cracking due to embrittlement).

ustry groups are in place whose objectives include the investigation of the aging effects licable to reactor vessel internals regarding such items as thermal or neutron irradiation rittlement (loss of fracture toughness), void swelling (change in dimensions), stress corrosion king (PWSCC and IASCC), and loss of pre-load for baffle and former-assembly bolts.

acceptance criteria for examinations performed in accordance with the Inservice Inspection gram: Reactor Vessel Internals are based on the applicable regulations and acceptance dards. Corrective actions for conditions that are adverse to quality are performed in ordance with the Corrective Action Program as part of the Quality Assurance Program. The ective action process provides reasonable assurance that deficiencies adverse to quality are er promptly corrected or are evaluated to be acceptable.

mmitments following action will be implemented al least two years prior to period of extended operation:

Reactor Vessel Internals Inspections Millstone will follow the industry efforts on reactor vessel internals regarding such issues as thermal or neutron irradiation embrittlement (loss of fracture toughness), void swelling (change in dimensions), stress corrosion cracking (PWSCC and IASCC), and loss of pre-load for baffle and former-assembly bolts and will implement the appropriate recommendations resulting from this guidance. The revised program description, including a comparison to the 10 program elements of the NUREG-1801 program, will be submitted to the NRC for approval.

This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 13.

following program enhancement will be implemented at least two years prior to the period of nded operation:

Augmented Holddown Spring Inspections Augmented inspection of the Millstone Unit 3 core barrel holddown spring will be performed. In particular, the inspection will detect gross indication of loss of preload as an aging effect. As an alternative to performing an augmented inspection, the holddown spring will be replaced. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 14.

28/18 19.2-13 Rev. 31

gram Description rvice Inspection Program: Systems, Components and Supports corresponds to the following REG-1801 program descriptions:

Section XI.M1, ASME Section XI Inservice Inspection, Subsection IWB, IWC, and IWD,Section XI.M3, Reactor Head Closure Studs,Section XI.M11, Ni-Alloy Nozzles and Penetrations,Section XI.M12, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS), and Section XI.S3, ASME Section XI, Subsection IWF.

Inservice Inspection Program: Systems, Components and Supports is an existing program was developed to comply with the requirements of ASME Boiler and Pressure Vessel Code, tion XI (Reference 19.2-11). The ASME program provides the requirements for ISI, repair, replacement for all Class 1, 2 and 3 components and the associated component supports. For nse renewal, the Millstone program has been credited to manage the effects of aging for only ss 1 and specific Class 2 components (on the secondary side of the steam generators as rmined through the aging management review process) and for Class 1, 2, and 3 components ports. Inservice Inspection Program: Systems, Components and Supports manages the aging cts of cracking, loss of fracture toughness, loss of material and loss of preload.

ustry programs are in place whose objectives include the investigation of aging effects licable to nickel-based alloys (i.e., PWSCC in Alloy 600 base metal and Alloy 82/182 weld als) and identification of appropriate aging management activities.

acceptance criteria for examinations performed in accordance with the Inservice Inspection gram: Systems, Components and Supports are based on the applicable regulations and eptance standards. Corrective actions for conditions that are adverse to quality are performed ccordance with the Corrective Action Program as part of the Quality Assurance Program. The ective action process provides reasonable assurance that deficiencies adverse to quality are er promptly corrected or are evaluated to be acceptable.

mmitments following action will be taken prior to the period of extended operation:

PWSCC of Nickel-Based Alloys 28/18 19.2-14 Rev. 31

and identifying the appropriate aging management activities and will implement the appropriate recommendations resulting from this guidance. The revised program description will be submitted at least two years prior to the period of extended operation for staff review and approval to determine if the program demonstrates the ability to manage the effects of aging in nickel based components per 10 CFR 50.54.21(a)(3). This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 15.

Monitoring Fracture Toughness For potentially susceptible CASS materials, either enhanced volumetric examinations or a unit or component specific flaw tolerance evaluation (considering reduced fracture toughness and unit specific geometry and stress information) will be used to demonstrate that the thermally-embrittled material has adequate fracture toughness in accordance with NUREG-1801 Section XI.M12.3. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 28.

Pressurizer Spray Head Assembly Cracking The pressurizer spray head assembly will be either replaced or inspected utilizing the best currently available (at the time of inspection) techniques for detecting cracking resulting from SCC. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 37.

.1.18 Inspection Activities: Load Handling Cranes and Devices gram Description ection Activities: Load Handling Cranes and Devices corresponds to NUREG-1801, Section M23, Inspection of Overhead Heavy Load (Related to Refueling) Handling Systems. The gram manages the aging effect of loss of material for the load handling cranes and devices hin the scope of license renewal. The in-scope load handling cranes and devices are either ty-related or seismically designed to ensure that they will not adversely impact safety-related ponents during or subsequent to a seismic event.

ection Activities: Load Handling Cranes and Devices addresses the overall condition of the e or device, including checking the condition of the structural members (i.e., rails, girders,

) and fasteners on the crane or device, the runways along which the crane or device moves, the baseplates and anchorages for the runways and monorails.

acceptance criterion for visual inspections is the absence of anomalous indications that are s of degradation. Corrective actions for conditions that are adverse to quality are performed in ordance with the Corrective Action Program as part of the Quality Assurance Program. The ective action process provides reasonable assurance that deficiencies adverse to quality are er promptly corrected or are evaluated to be acceptable.

28/18 19.2-15 Rev. 31

following program enhancements will be implemented prior to the period of extended ration:

Inclusion of In-Scope Lifting Devices The existing inspection program will be modified to include those lifting devices that require monitoring for license renewal, but are not already included in the program. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 16.

Inspection Criteria Implementing procedures and documentation will be modified to include visual inspections for the loss of material on the crane and trolley structural components and the rails in the scope of license renewal. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 17.

.1.19 Reactor Vessel Surveillance gram Description ctor Vessel Surveillance corresponds to NUREG-1801,Section XI.M31 Reactor Vessel veillance. The Reactor Vessel Surveillance program manages the aging effect of loss of ture toughness due to neutron embrittlement of the low-alloy subcomponents in the beltline on of the reactor vessel. Neutron dosimetry and material properties data derived from the tor vessel materials irradiation surveillance program are used in calculations and evaluations demonstrate compliance with applicable regulations. This program ensures compliance with hnical Requirements Manual requirements that surveillance specimens are removed and mined at predetermined intervals established in the Technical Specification to monitor the nges in the material properties and the results of the examinations used to update the Technical cification operating limits.

acceptance criteria are established in the current licensing basis as compliance with the licable regulations and standards. Corrective actions for conditions that are adverse to quality performed in accordance with the Corrective Action Program as part of the Quality Assurance gram. The corrective action process provides reasonable assurance that deficiencies adverse to lity are either promptly corrected or are evaluated to be acceptable.

.1.20 Service Water System (Open-Cycle Cooling) gram Description Service Water System (Open-Cycle Cooling) program corresponds to NUREG-1801, Section M20, Open Cycle Cooling Water System. The program manages the aging effects of loss of erial and buildup of deposits. The program implements the NRC guidelines in Generic 28/18 19.2-16 Rev. 31

nsure that corrosion (including microbiologically influenced corrosion), erosion, protective ting failure, silting, and biofouling do not degrade the performance of safety-related systems iced by Service Water System; (d) a system walkdown inspection to ensure compliance with licensing basis; and (e) a review of maintenance, operating, and training practices and cedures. Millstone Unit 3 relies on either frequent, regular inspection and cleaning of heat hangers, thermal performance testing of heat exchangers, or maintaining of heat exchangers in lay-up to preclude fouling.

acceptance criterion for visual inspections is the absence of anomalous indications that are s of degradation. Corrective actions for conditions that are adverse to quality are performed in ordance with the Corrective Action Program as part of the Quality Assurance Program. The ective action process provides reasonable assurance that deficiencies adverse to quality are er promptly corrected or are evaluated to be acceptable.

.1.21 Steam Generator Structural Integrity gram Description m Generator Structural Integrity corresponds to NUREG-1801,Section XI.M19, Steam erator Tube Integrity Program. This program manages the aging effects of loss of material cracking and adopts the performance criteria and guidance for monitoring and maintaining m generator tubes as defined in NEI 97-06 (Reference 19.2-13). The program incorporates ormance criteria for structural integrity, accident-induced leakage, and operational leakage.

program includes preventive measures to mitigate degradation through the control of primary secondary side water chemistry; assessment of degradation mechanisms; inservice inspection he steam generator tubes to detect degradation; evaluation and plugging or repair, as needed; leakage monitoring to ensure the structural and leakage integrity of the pressure boundary.

ustry programs are in place whose objectives include the investigation of aging effects licable to nickel-based alloys (i.e., PWSCC in Alloy 600 base metal and Alloy 82/182 weld als) and identification of appropriate aging management activities.

acceptance criteria are established in the current licensing basis as compliance with the licable regulations and acceptance standards. Corrective actions for conditions that are erse to quality are performed in accordance with the Corrective Action Program as part of the lity Assurance Program. The corrective action process provides reasonable assurance that ciencies adverse to quality are either promptly corrected or are evaluated to be acceptable.

mmitments following action will be implemented based on the availability of the industry guidance:

PWSCC of Nickel-Based Alloys 28/18 19.2-17 Rev. 31

and identifying the appropriate aging management activities and will implement the appropriate recommendations resulting from this guidance. The revised program description will be submitted prior to the period of extended operation for staff review and approval to determine if the program demonstrates the ability to manage the effects of aging in nickel based components per 10 CFR 50.54.21(a)(3).

This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 15.

.1.22 Structures Monitoring Program gram Description ctures Monitoring Program corresponds to the following NUREG-1801 program criptions:

Section XI.S5 Masonry Wall Program,Section XI.S6 Structures Monitoring Program, and Section XI.S7 R.G. 1.127, Inspection of Water Control Structures Associated with Nuclear Power Plants.

Structures Monitoring Program manages the aging effects of loss of material, change of erial properties, and cracking by the monitoring of structures and structural support systems are in the scope of license renewal. The majority of these structures and structural support ems are monitored under 10 CFR 50.65 (Reference 19.2-14). Other structures in the scope of nse renewal (such as non-safety related buildings and enclosures, duct banks, valve pits and ches, HELB barriers, and flood gates) are also monitored to ensure there is no loss of intended ction.

scope includes all masonry walls and water-control structures identified as performing nded functions in accordance with 10 CFR 54.4.

acceptance criterion for visual inspections is the absence of anomalous indications that are s of degradation. Corrective actions for conditions that are adverse to quality are performed in ordance with the Corrective Action Program as part of the Quality Assurance Program. The ective action process provides reasonable assurance that deficiencies adverse to quality are er promptly corrected or are evaluated to be acceptable.

mmitments following program enhancements will be implemented prior to the period of extended ration:

28/18 19.2-18 Rev. 31

NUREG-1801 recommends the use of ACI 349.3R-96 and ANSI/ASCE 11-90, as a reference for recommendations for the development of an evaluation procedure for nuclear safety-related concrete structures and existing buildings. These documents were not used or referenced as a standard for establishing the Structures Monitoring Program.

The implementing procedures will be modified to include ACI 349.3R-96 and ANSI/ASCE 11-90 as references and as input documents for the inspection program. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 18.

Addition of Structures to the Structures Monitoring Program The Structures Monitoring Program does not currently monitor all structures in-scope for license renewal. The Structures Monitoring Program and implementing procedures will be modified to include all in-scope structures. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 19.

Sampling of Groundwater Groundwater samples will be taken on a periodic basis, considering seasonal variations, to ensure that the groundwater is not sufficiently aggressive to cause the below-grade concrete to degrade. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 20.

Engineering Notification of Submerged Medium Voltage Cables The Structures Monitoring Program and implementing procedures will be modified to alert the appropriate engineering organization if the structures inspections identify that medium voltage cables in the scope of license renewal have been submerged. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 21.

Inspection of Normally Inaccessible Areas That Become Accessible The maintenance and work control procedures will be revised to ensure that inspections of inaccessible areas are performed when the areas become accessible by such means as excavation or installation of shielding during maintenance or for any other reason. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 22.

.1.23 Tank Inspection Program gram Description k Inspection Program corresponds to NUREG-1801,Section XI.M29, Aboveground Carbon l Tanks. The program manages the aging effect of loss of material through periodic internal external tank inspections. The program includes inspections of the sealant and caulking in and und the tank and the concrete foundation and evaluations to monitor the condition of coatings, 28/18 19.2-19 Rev. 31

aces of tank bottoms.

acceptance criterion for visual inspections of paint, coatings, sealant, caulking, and structural ments is the absence of anomalous indications that are signs of degradation. Thickness surements of the tank walls and bottoms are evaluated against design thickness, established eline values, or loss of material allowances. Corrective actions for conditions that are adverse uality are performed in accordance with the Corrective Action Program as part of the Quality urance Program. The corrective action process provides reasonable assurance that deficiencies erse to quality are either promptly corrected or are evaluated to be acceptable.

mmitments following program enhancements will be implemented prior to the period of extended ration:

Inspection of sealants and caulkings Appropriate inspections of sealants and caulkings used for moisture intrusion prevention in and around aboveground tanks will be performed. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 23.

Non-destructive Volumetric Examination of Inaccessible Tank Bottoms Non-destructive volumetric examination of the in-scope inaccessible locations, such as the external surfaces of tank bottoms, will be performed prior to the period of extended operation. Subsequent inspections will be performed on a frequency consistent with scheduled tank internals inspection activities. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 24.

Tanks Being Added to Tank Inspection Program The security diesel fuel oil tank and diesel fire pump fuel oil tank are in-scope for license renewal and have been included in the respective Tank Inspection Program inspection plan. These changes complete the action required for commitment Item 25 in Table 19.6-1.

.1.24 Work Control Process gram Description rk Control Process is a plant specific program that integrates and coordinates the combined rts of Maintenance, Engineering, Operations, and other support organizations to manage ntenance activities. The Work Control Process is utilized to manage the aging effects of loss of erial, change of material properties, cracking, and buildup of deposits for components and 28/18 19.2-20 Rev. 31

trol Process. The Work Control Process also provides opportunities to collect oil and engine lant fluid samples for subsequent analysis of contaminants and chemical properties, which ld either indicate or affect aging.

ddition to visual inspections, the field inspection for loss of material due to selective leaching include mechanical means, such as resonance when struck by another object, scraping, or ping.

acceptance criterion for visual inspections is the absence of anomalous signs of degradation.

acceptance criteria for testing or sampling are specified in the various station procedures

/or vendor technical manuals or recommendations. Corrective actions for conditions that are erse to quality are performed in accordance with the Corrective Action Program as part of the lity Assurance Program. The corrective action process provides reasonable assurance that ciencies adverse to quality are either promptly corrected or are evaluated to be acceptable.

mmitments following program enhancements will be implemented prior to the period of extended ration:

Performance of Inspections During Maintenance Activities Changes will be made to maintenance and work control procedures to ensure that inspections of plant components and plant commodities will be appropriately and consistently performed and documented for aging effects during maintenance activities.

This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 26.

Selective Leaching Inspection Using the Work Control Process, a baseline inspection for the loss of material due to selective leaching will be performed on a representative sample of locations for susceptible materials by visual, and mechanical or other appropriate methods prior to entering the period of extended operation. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 31.

Verification of Program Scope A review of the Work Control Process inspection opportunities for each material and environment group supplemental to the initial review, supplemental to the initial review conducted during the development of the LRA, will be performed. Baseline inspections will be performed for the material and environment combinations that have not been inspected as part of the Work control Process. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 32.

28/18 19.2-21 Rev. 31

gram Description Bolting Integrity Program corresponds to NUREG-1801,Section XI.M18, Bolting grity. The program manages the aging effects of cracking, loss of material and loss of oad.

s is accomplished by establishing good bolting practices in accordance with EPRI NP-5067, d Bolting Practices, A Reference Manual for Nuclear Power Plant Maintenance Personnel, ume 1: Large Bolt Manual, and Volume 2: Small Bolts and Threaded Fasteners and EPRI 104213, Bolted Joint Maintenance and Application Guide. For ASME Class bolting, aging cts are additionally managed by the performance of inservice examinations in accordance h ASME Section XI, Subsections IWB, IWC, IWD, and IWF.

engineering evaluations determine if a component needs to be repaired/replaced or is eptable for continued operation until the next scheduled inspection. Corrective actions for ditions that are adverse to quality are performed in accordance with the Corrective Action gram as part of the Quality Assurance Program. The corrective action process provides onable assurance that deficiencies adverse to quality are either promptly corrected or are luated to be acceptable.

.2 REFERENCES FOR SECTION 19.2

-1 TR-105714, PWR Primary Water Chemistry Guidelines, Technical Report, Revision 3, Electric Power Research Institute.

-2 TR-102134, PWR Secondary Water Chemistry Guidelines, Technical Report, Revision 3, Electrical Power Research Institute.

-3 EPRI TR-107396, Closed Cooling Water Chemistry Guideline, Technical Report, Electrical Power Research Institute, Palo Alto, CA, November 1997.

-4 NRC Interim Staff Guidance (ISG)-05, The Identification And Treatment of Electrical Fuse Holders For License Renewal, U.S. Nuclear Regulatory Commission, March 10, 2003.

-5 Letter from Pao-Tsin Kuo, Nuclear Regulatory Commission, to Alex Marion, Nuclear Energy Institute, and David Lochbaum, Union of Concerned Scientists, Proposed Interim Staff Guidance (ISG)-15: Revision of Generic Aging Lessons Learned (GALL)

Aging Management Program (AMP) X1.E2, Electrical Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits, August 12, 2003.

-6 NRC Interim Staff Guidance (ISG)-04, Aging Management of Fire Protection Systems for License Renewal, U.S. Nuclear Regulatory Commission, December 3, 2002.

28/18 19.2-22 Rev. 31

-8 ASTM D 1796, Standard Test Method for Water and Sediment in Fuel Oils by the Centrifuge Method, American Society for Testing Materials, West Conshohocken, PA.

-9 ASTM D 2276, Standard Test Method for Particulate Contaminant in Aviation Fuel by Line Sampling, American Society for Testing Materials, West Conshohocken, PA.

-10 ASTM D 4057, Standard Practice for Manual Sampling of Petroleum and Petroleum Products, American Society for Testing Materials, West Conshohocken, PA.

-11 ASME Boiler and Pressure Vessel Code Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, American Society of Mechanical Engineers.

2-12 Generic Letter 89-13, Service Water System Problems Affecting Safety-Related Equipment, Nuclear Regulatory Commission, July 18, 1989 (Supplement 1 dated 4/4/90).

-13 NEI 97-06, Steam Generator Program Guidelines, Technical Report, Nuclear Energy Institute.

-14 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, U. S. Nuclear Regulatory Commission.

-15 NRC Bulletin 2001-01, Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles, U.S. Nuclear Regulatory Commission, August 3, 2001.

2-16 NRC Bulletin 2002-01, Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity, U.S. Nuclear Regulatory Commission, March 18, 2002.

-17 NRC Bulletin 2002-02, Reactor Pressure Vessel Head and Vessel Head Penetration Nozzle Inspection Programs, U.S. Nuclear Regulatory Commission, August 9, 2002.

2-18 NRC Bulletin 2003-02, Leakage from Reactor Pressure Vessel Lower Head Penetrations and Reactor Coolant Pressure Boundary Integrity, U.S. Nuclear Regulatory Commission, 08/21/03.

-19 NRC Order EA-03-009, Issuance of Order Establishing Interim Inspection Requirements For Reactor Pressure Vessel Heads At Pressurized Water Reactors, U.S. Nuclear Regulatory Commission, February 11, 2003.

2-20 NRC Bulletin 2004-01, Inspection of Alloy 82/182/600 Materials used in the Fabrication of Pressurizer Penetrations and Steam Space Piping Connections at Pressurized-Water Reactors, May 28, 2004.

28/18 19.2-23 Rev. 31

part of the application for a renewed license, 10 CFR 54.21(c) requires that an evaluation of e-limited Aging Analyses (TLAAs) for the period of extended operation be provided. The owing TLAAs have been identified and evaluated to meet this requirement.

.1 REACTOR VESSEL NEUTRON EMBRITTLEMENT reactor vessel is described in FSAR Section 5.3. Time-limited aging analyses (TLAAs) licable to the reactor vessel are:

Upper-shelf energy.

Pressurized thermal shock.

Pressure-temperature limits.

Reactor Vessel Surveillance program manages reactor vessel irradiation embrittlement zing subprograms to monitor, calculate, and evaluate the time-dependent parameters used in aging analyses for pressurized thermal shock, upper-shelf energy, and pressure-temperature t curves to ensure continuing vessel integrity through the period of extended operation.

reactor vessel neutron embrittlement evaluations have been based on 54 effective full power rs of operation. 54 effective full power years would be reached at the end of the period of nded operation (60 years) assuming a capacity factor of 90% for the lifetime of the unit.

.1.1 Upper Shelf Energy CFR 50, Appendix G contains screening criteria that establish limits on how far the upper shelf rgy values for a reactor pressure vessel material may be allowed to drop due to neutron diation exposure. The regulation requires the initial upper shelf energy value to be greater than t-lbs in the unirradiated condition and for the value to be greater than 50 ft-lbs in the fully diated condition as determined by Charpy V-notch specimen testing throughout the licensed of the plant. Upper shelf energy values of less than 50 ft-lbs may be acceptable to the NRC if n be demonstrated that these lower values will provide margins of safety against brittle ture equivalent to those required by ASME Section XI, Appendix G.

eptable upper shelf energy values have been calculated in accordance with Regulatory Guide

, Revision 2 to the end of the period of extended operation. Calculated upper shelf energy es for the most limiting reactor pressure vessel beltline plate and weld materials remain ter than 50 ft-lbs.

.1.2 Pressurized Thermal Shock ctor pressure vessel beltline fluence is one of the factors used to determine the margin to tor pressure vessel pressurized thermal shock as a result of radiation embrittlement. The 28/18 19.3-1 Rev. 31

line material (RTPTS) and the screening criteria established in accordance with CFR 50.61(b)(2). The screening criteria for the limiting reactor vessel materials are 270°F for line plates, forging and axial weld materials, and 300°F for beltline circumferential weld erials.

eptable RTPTS values have been calculated in accordance with Regulatory Guide 1.99, ision 2, requirements to the end of the period of extended operation.

.1.3 Pressure-Temperature Limits CFR Part 50 Appendix G requires that heatup and cooldown of the reactor pressure vessel be omplished within established pressure-temperature limits. These limits identify the maximum wable pressure as a function of reactor coolant temperature. As the pressure vessel becomes diated and its fracture toughness is reduced, the allowable pressure at low temperatures is uced. Therefore, in order to heatup and cooldown the Reactor Coolant System, the reactor lant temperature and pressure must be maintained within the limits of Appendix G as defined he reactor vessel fluence.

tup and cooldown limit curves have been calculated using the adjusted RTNDT corresponding he limiting beltline material of the reactor pressure vessel for the current period of licensed ration. Current cold overpressure protection system (COPS) heatup and cooldown limit curves e approved in License Amendment 197.

ccordance with 10 CFR 50, Appendix G, updated pressure-temperature limits for entering the od of extended operation will be developed and implemented prior to the period of extended ration. Cold overpressure protection system enable temperature requirements will be updated nsure that the pressure-temperature limits will not be exceeded for postulated plant transients ng the period of extended operation. Millstone Unit 3 will calculate USE, RTPTS, and P-T ts based on fluence values developed in accordance with Regulatory Guide 1.190 uirements, as amended or superseded by future regulatory guidance changes, through the od of extended operation.

.2 METAL FATIGUE gue is defined as structural deterioration that can occur through repeated stress or strain cycles lting from fluctuations in loads and/or temperatures. After repeated cyclic loading of icient magnitude, micro-structural damage can accumulate leading to microscopic crack ation at the most highly affected locations. Fatigue cracks typically initiate at points of imum local stress ranges and minimum local strength. Further cyclic mechanical and/or mal loading can lead to crack growth.

gue represents an aging mechanism. As such, fatigue evaluations represent a time-limited g analysis even though the system, structure and component design limits are based upon the 28/18 19.3-2 Rev. 31

3.2.1 Millstone Unit 3 Class 1 Components mponents within the Millstone Unit 3 nuclear steam supply system are subject to a wide ety of varying mechanical and thermal loads that contribute to fatigue accumulation. The ctor Coolant System components are designed in accordance with ASME Boiler and Pressure sel Code,Section III (Reference 19.3-21) this code requires that design analyses for Class 1 ems and components address fatigue and the establishment of load limits to preclude initiation atigue cracks.

type and number of Reactor Coolant System design transients have been identified. In all ances, the number of Reactor Coolant System design transients assumed in the original design e found to be acceptable for the period of extended operation.

C Bulletin 88-08 identified a concern regarding potential temperature stratification or perature oscillations in unisolable sections of piping attached to the Reactor Coolant System.

ed upon the Millstone Unit 3 response (Reference 19.3-22) and supplemental munications, the NRC concluded that Millstone Unit 3 meets the requirements of Bulletin 88-Reference 19.3-23).

ssurizer surge line thermal stratification was a concern raised by the NRC in Bulletin 88-11.

of the requirements of this bulletin was to analyze the effects of thermal stratification on e line integrity. These analyses were collectively performed as a Westinghouse Owners Group (Reference 19.3-24) supplemented by additional unit specific inspections and activities.

ed upon the Westinghouse Owners Group task, the NRC concluded that the bounding luations and supplemental unit specific inspections and activities demonstrate that the lstone Unit 3 pressurizer surge line piping and associated nozzles meet Bulletin 88-11 uirements (References 19.3-25, and 19.3-26, and 19.3-27). The NRC has reviewed this rmation and determined that Millstone Unit 3 has addressed the actions required by Bulletin 11 (Reference 19.3-28).

rmal aging refers to changes in the microstructure and properties of a susceptible material due rolonged exposure to elevated temperatures above approximately 480°F. Reactor Coolant tem temperatures exceed this threshold. At these temperatures, the hardness of potentially eptible Cast Austenitic Stainless Steel (CASS) materials increase while their ductility, impact ngth and more importantly, their fracture toughness, decrease. Fracture toughness is one of the e important design inputs in a leak-before-break and a flaw tolerance evaluation, performed to ure protection of the reactor coolant system against guillotine pipe breaks throughout plant The degree of change in fracture toughness (thermal embrittlement) is dependent on the time xposure to these elevated temperatures.

ept for the pressurizer spray head, acceptable thermal and pressure transients, and operating les have been projected for ASME Section III, Class 1 components, through the period of 28/18 19.3-3 Rev. 31

mmitments following actions will be implemented prior to the period of extended operation:

rmal aging of the pressurizer spray head will be managed by the Inservice Inspection gram: Systems, Components and Supports. This commitment is identified in Table 19.6-1, ense Renewal Commitments, Item 28.

.2.2 Non-Class 1 Components

-Class 1 components can include ASME Section III Classes 2 and 3, ANSI Standard B31.7 sses 2 and 3, and ANSI Standard B31.1 (Reference 19.3-29) piping and tubing. Piping systems gned to these requirements (e.g., sample lines) incorporate a stress range reduction factor to servatively address the effects of thermal cycling on fatigue. For those sample lines projected xperience greater than 7,000 equivalent full-temperature thermal cycles, actual expansion sses did not exceed allowable expansion stresses.

eptable numbers of thermal cycles and acceptable expansion stresses have been projected to end of the period of extended operation.

.2.3 Environmentally Assisted Fatigue effect of reactor coolant environment on fatigue is generally referred to as environmentally sted fatigue. As part of an industry effort to address environmental effects on operating lear power plants during the current 40-year licensing term, Idaho National Engineering oratories evaluated fatigue-sensitive component locations at plants designed by all four estic nuclear steam supply system vendors. These evaluations are presented and discussed in REG/CR-6260 (Reference 19.3-30). The evaluations associated with the newer-vintage tinghouse plants are applicable, since the majority of the Millstone Unit 3 Class 1 systems components were designed to ASME Section III requirements.

influence of the reactor water environment on the cumulative usage factor was evaluated for following representative components identified in NUREG/CR-6260 for the period of nded operation, using the most recent laboratory data and methods:

Reactor vessel shell and lower head.

Reactor vessel inlet and outlet nozzles.

Surge line.

Charging nozzle.

28/18 19.3-4 Rev. 31

Residual Heat Removal System Class 1 piping.

se six fatigue-sensitive locations have been evaluated using the methods identified in REG/CR-6583 (Reference 19.3-31), and NUREG/CR-5704 (Reference 19.3-32).

izing Millstone Unit 3 cyclic and transient information, four fatigue sensitive component tions were determined to have cumulative usage factors (CUFs) greater than 1.0 over the od of extended operation. For the pressurizer surge line, charging nozzle, safety injection zles, and Residual Heat Removal System piping, more detailed stress analyses or fatigue nitoring and cycle counting would have to be used to reduce CUF below 1.0. Due to code servatisms included in the ASME Code, a CUF of greater than 1.0 does not mean that fatigue king will occur; only that there is a potential for fatigue cracking to occur over the period of nded operation. Utilizing these conservatisms, an approach will be developed to manage the cts of environmentally assisted fatigue for those specific locations with a CUF greater than The expected approach is to manage these effects through the use of an inspection program has been reviewed and approved by the NRC. The program would be expected to include, for mple, appropriate non-destructive examinations and NRC acceptable inspection periods.

air or replacement activities would be based upon inspection results.

mmitments following actions will be implemented prior to the period of extended operation:

The effects of environmentally assisted fatigue for those specific locations with a CUF greater than 1.0 will be managed by the Metal Fatigue of Reactor Coolant Pressure Boundary program for the period of extended operation. If the specific locations are not repaired, replaced, or successfully re-analyzed, a modified inspection program description, including a comparison to the 10 program elements of NUREG-1801 program, will submitted to the NRC for approval. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 27.

Millstone will follow industry efforts that will provide specific guidance to license renewal applicants for evaluating the environmental effects of fatigue on applicable locations, other than those identified in NUREG/CR-6260. Millstone will also implement the appropriate recommendations resulting from this guidance. Until these recommendations are available, Millstone 3 commits to using the pressurizer surge line nozzle as a leading indicator to address environmental effects on fatigue of pressurizer sub-components during the period of extended operation. This commitment is identified in Table 19.6-1, License Renewal Commitments, Item 29.

.3 ENVIRONMENTAL QUALIFICATION (EQ) OF ELECTRIC EQUIPMENT ctrical Equipment Qualification (EEQ) program is an integral part of the design, construction operation of nuclear power generating stations. A description of this program provided in 28/18 19.3-5 Rev. 31

CFR Part 50 requires that certain categories of systems, structures and components be gned to accommodate the effects of both normal and accident environmental conditions, and design control measures be employed to ensure the adequacy of these designs. Specific uirements pertaining to the environmental qualification of these categories of electrical ipment are embodied within 10 CFR 50.49 (Reference 19.3-33). The categories include ty-related (Class 1E) electrical equipment, non-safety-related electrical equipment whose ure could prevent the satisfactory accomplishment of a safety function by safety-related ipment, and certain post-accident monitoring equipment. As required by 10 CFR 50.49, trical equipment not qualified for the current license term is to be refurbished, replaced or e its qualification extended prior to reaching the aging limits established in the evaluation.

ng evaluations for electrical equipment that specify a qualification of 40 years or greater are sidered to represent a time-limited aging analysis. Guidance relating to the methods and cedures for implementing the requirements of 10 CFR 50.49 is contained within Regulatory de 1.89 (Reference 19.3-34). Further guidance for post-accident monitoring equipment is tained within Regulatory Guide 1.97 (Reference 19.3-35).

ironmental qualification of electrical equipment will be adequately managed for the period of nded operation.

.4 CONTAINMENT LINER PLATE, METAL CONTAINMENTS, AND PENETRATIONS FATIGUE ANALYSIS 3.4.1 Containment Liner Plate lstone Unit 3 has a conventionally reinforced concrete Containment structure maintained at atmospheric pressure, surrounded by an enclosure building. A welded carbon steel liner plate tached to the inside surface of the concrete, providing a high degree of leak tightness.

mponents of the liner plate include penetration sleeves, access openings, and piping etrations.

luations of the Containment liner plate involve the use of time-limited assumptions such as osion rates and thermal cycles. These evaluations meet the requirements of 10 CFR 54.3 and, uch, represent time-limited aging analyses. Acceptable Containment liner plate integrity has n projected to the end of the period of extended operation.

.4.2 Containment Penetrations lstone Unit 3 Containment penetrations are used for personnel and equipment access, process ng, electrical service, or for a mechanical fuel transfer system. Each of these penetrations is hored to, and transfers loads to the reinforced Containment wall. There were no applicable es for the design of concrete Containment liners at the beginning of the construction of the lstone Unit 3 liner. ASME Section III, Division 1 and 2, and ASME Section VIII were used as des.

28/18 19.3-6 Rev. 31

CFR 54.3 and, as such, represent time-limited aging analyses.

eptable Containment penetration integrity has been projected to the period of extended ration.

.5 OTHER PLANT-SPECIFIC TIME-LIMITED AGING ANALYSES

.5.1 Crane Load Cycle Limit containment polar crane, spent fuel crane, monorails, and jib cranes are examples of the types ranes within the scope of license renewal. These cranes meet the guidance contained in REG-0612.

evaluation of crane loads represents a time-limited aging analysis per 10 CFR 54.3 since it olves the use of a time-limited assumption, load cycles. The most frequently used crane is the nt fuel crane. Considering all uses, the spent fuel crane is expected to conservatively erience a total number of load cycles over a 60-year period, that is well below the number of les allowed in Crane Manufacturers Association of America, Inc. Specification No. 70.

eptable crane load cycles have been projected to the end of the period of extended operation.

3.5.2 Reactor Coolant Pump Flywheel reactor coolant pump motors are provided with flywheels to increase rotational inertia, thus onging pump coast-down and assuring a more gradual loss of primary coolant flow to the core he event that pump power is lost. During normal operation, the reactor coolant pump flywheels elop sufficient kinetic energy to produce high-energy missiles in the event of failure.

ditions that may result in overspeed of the pump increase both the potential for failure and the tic energy of the flywheel.

stinghouse Report WCAP-14535A, Topical Report on Reactor Coolant Pump Flywheel ection Elimination (Reference 19.3-36) presents an evaluation of the likelihood of flywheel ure over a 60-year period of operation and the justification for relaxation of RG 1.14, Revision egulatory Position C.4.b(1), requirements to those identified in Regulatory Position C.4.b(2).

ng this evaluation, the NRC issued Amendment No. 169 to the unit Technical Specifications, sistent with RG 1.14, Revision 1, Regulatory Position C.4.b(2), to allow the examination of h reactor coolant pump flywheel at least once every 10-years, coinciding with the ASME tion XI inservice inspection program schedule.

evaluation of reactor coolant pump flywheels represents a time-limited aging analysis per CFR 54.3 since it involves the use of time limited assumptions such as thermal cycles and k growth rates. This evaluation, which indicates a low likelihood of flywheel fatigue failure r a 60-year period, along with implementation of the Inservice Inspection Program: Systems, 28/18 19.3-7 Rev. 31

ctor coolant pump flywheel fatigue cracking will be adequately managed for the period of nded operation.

.5.3 Leak-Before-Break Leak-Before-Break (LBB) analysis was evaluated as a time-limited aging analysis (TLAA) etermine that the analysis remains valid for the period of extended operation. The reactor lant system loop piping (hot leg, cold leg and crossover piping) has been evaluated for LBB.

LBB analysis was determined to remain valid for the period of extended operation by luating their time-based inputs. Thermal aging of cast austenitic stainless steel (CASS) erials and fatigue crack growth calculations were determined to be time-based inputs as ned in 10 CFR 54.3 and required evaluation for the period of extended operation.

metal fatigue TLAA evaluations described in FSAR Section 19.3.2.1 conclude that design s limits are not exceeded for ASME Class 1 components (which envelopes the components luated for LBB) through the period of extended operation.

rmal aging of CASS materials for components that have been evaluated for LBB has been luated as a TLAA since long-term exposure of CASS materials to reactor coolant system rating temperatures results in an increase in material hardness while its ductility, impact ngth and fracture toughness decrease. Fracture toughness represents one of the more important gn inputs in a LBB evaluation. The degree of reduction in CASS fracture toughness is endent on the time of thermal exposure. However, the change in material properties due to mal aging reaches a saturation value, after which material property changes resulting from itional thermal exposure are not significant. The evaluation of the thermal aging of CASS erial for the LBB evaluations consisted of a review to determine whether the fracture ghness value used in the analysis was conservative relative to the fully aged value for fracture ghness for the CASS components. The review concluded that the analysis values were either al to or lower than the worst-case saturation (fully aged) values for fracture toughness in all

s. Therefore, since the CASS material property values used in current design basis LBB luations represent fully aged (saturation) values, and since these values would not change with her exposure time, the LBB evaluations remain valid for the thermal aging of CASS materials ughout the period of extended operation.

LBB analysis has been projected to remain valid through the end of the period of extended ration.

.5.4 Containment Subfoundation Unit 3 Containment basemat is 10 feet thick and is supported by a subfoundation, which is nded on bedrock. The subfoundation consists of (from bottom to top): (1) a 10 inch layer of ous concrete made of Portland cement and coarse aggregate, (2) approximately 1/16-inch 28/18 19.3-8 Rev. 31

mina Cement, or HAC layer], and (5) thin mortar seal.

987, Unit 3 identified cement constituents (calcium-alumina, which forms a white residue) in drainage system installed in the HAC layer of the Containment subfoundation. An evaluation rmined that the rubber waterproofing membrane had developed leaks, which allowed for the ess of water into the HAC layer.

e tests and plate bearing tests were conducted, along with additional testing on HAC mock-that were built to the same specifications as used in the original construction of the MP3 foundation.

eral core samples were removed from the HAC porous concrete layer in the subfoundation of ESF Building, where a portion of the building subfoundation is the same as that for the tainment basemat. Tests were conducted on these samples to quantify the available margin in bearing stresses below the containment basemat, for the current license period of 40 years.

005, a condition assessment was performed to determine the acceptability of the Unit 3 tainment subfoundation porous concrete layers for the period of extended operation.

mputation of bearing stresses on the porous cement surface showed that for a bounding loss of n as much as 7.4% in foundation area (or volume), the bearing stress remains significantly less the tested strength of 2850 psi. The amount of loss in this scenario bounds the projection of total amount of white residue that is conservatively calculated to be collected from the struction of the plant through the period of extended operation.

evaluation of the Millstone Unit 3 containment subfoundation represents a time-limited aging lysis per 10 CFR 54.3 since it involves the use of time limited assumptions such as the imum amount of calcium-alumina that can be leached over time from the HAC layer and still ntain adequate support for the containment basemat.

structural integrity of the Millstone 3 (MP3) Containment subfoundation has been onstrated through the period of extended operation. Consistent with 10 CFR 54.21(c)(1),

ion (ii), the analyses have been projected to the end of the period of extended operation.

3.6 REFERENCES

FOR SECTION 19.3

-21 ASME Section III, Rules for Construction of Nuclear Vessels, ASME Boiler and Vessel Pressure Code, American Society of Mechanical Engineers, 1971.

-22 Letter from E. J. Mroczka to NRC, Response to NRC Bulletin No. 88-08, Thermal Stresses in Piping Connected to Reactor Coolant System, September 20, 1988.

-23 Letter from D. H. Jaffe to E. J. Mroczka, NRC Bulletin 88- 08, Thermal Stresses in Piping Connected to Reactor Coolant Systems (TAC No.69636, 69651 and 69653),

September 25, 1991.

28/18 19.3-9 Rev. 31

-25 Letter from J. F. Stolz to E. J. Mroczka, NRC Bulletin No. 88-11, Pressurizer Surge Line Thermal Stratification - Evaluation of Westinghouse Owners Group Bounding Analysis (TAC No. 72136 and 72145), August 6, 1990.

-26 Letter from J.F. Stolz to E. J. Mroczka, Pressurizer Surge Line Thermal Stratification, Bulletin 88-11, Millstone Unit 3 and Haddam Neck (TAC No. 72145 and 72136), July 31, 1991.

-27 Letter from J. F. Opeka to NRC, NRC Bulletin 88 Pressurizer Surge Line Thermal Stratification Final Submittal of Plant-Specific Reports, May 1, 1992.

-28 Letter from J. F. Stolz to J. F. Opeka, Response to NRC Bulletin No. 88 Pressurizer Surge Line Thermal Stratification for Haddam Neck Plant (TAC No. M72136) and Millstone 3 (TAC No. M72145), July 9, 1992.

-29 ANSI B31.1, Power Piping Code, American Society of Mechanical Engineers, 1967.

-30 NUREG/CR-6260, Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components, U.S. Nuclear Regulatory Commission.

-31 NUREG/CR-6583, Effects of LWR Coolant Environments on Fatigue Design Curves of Carbon and Low-Alloy Steels, U.S. Nuclear Regulatory Commission.

-32 NUREG/CR-5704, Effects of LWR Coolant Environment on Fatigue Design Curves of Austenitic Stainless Steel, U.S. Nuclear Regulatory Commission.

-33 10 CFR 50.49, Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants, U. S. Nuclear Regulatory Commission.

-34 Regulatory Guide 1.89, Environmental Qualification of Certain Electrical Equipment Important to Safety for Nuclear Power Plants, U. S. Nuclear Regulatory Commission.

-35 Regulatory Guide 1.97, Instrumentation of Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident, U. S. Nuclear Regulatory Commission.

3-36 WCAP-14535A, Topical Report on Reactor Coolant Pump Flywheel Inspection Elimination, Westinghouse Energy Systems, November 1996.

28/18 19.3-10 Rev. 31

.1 ELECTRICAL EQUIPMENT QUALIFICATION gram Description Electrical Equipment Qualification program corresponds to the Time-Limited Aging lysis (TLAA) support program described in NUREG-1801,Section X.E1, Environmental lification (EQ) of Electrical Components. The program applies to certain electrical ponents that are important to safety and could be exposed to post-accident environmental ditions, as defined in 10 CFR 50.49. The EEQ program ensures the continued qualification of equipment during and following design basis accidents. The program determines the essity for, and frequency of, component replacement or refurbishment in order to maintain the lification of the equipment. Performance of preventive maintenance and surveillance vities, and monitoring of normal ambient conditions, ensure that components remain within bounds of their original qualification and provide a basis for extending qualified life through nalysis.

acceptance criterion is that the equipment remains within the bounds of its qualified life such after maximum normal service conditions, the equipment retains sufficient capacity to orm its required safety function during design basis accident conditions. Corrective actions conditions that are adverse to quality are performed in accordance with the Corrective Action gram as part of the Quality Assurance Program. The corrective action process provides onable assurance that deficiencies adverse to quality are either promptly corrected or are luated to be acceptable.

.2 METAL FATIGUE OF REACTOR COOLANT PRESSURE BOUNDARY gram Description Metal Fatigue of Reactor Coolant Pressure Boundary program mitigates fatigue cracking sed by cyclic strains in metal components of the reactor coolant pressure boundary. This is omplished by monitoring and tracking the number of critical thermal and pressure transients selected Reactor Coolant System components to ensure that the number of design transient les is not exceeded during the plant operating life.

acceptance criterion is the fatigue usage factors bounded by the design usage factors.

rective actions for conditions that are adverse to quality are performed in accordance with the rective Action Program as part of the Quality Assurance Program. The corrective action cess provides reasonable assurance that deficiencies adverse to quality are either promptly ected or are evaluated to be acceptable.

28/18 19.4-1 Rev. 31

requirements of 10 CFR 54.21(c) stipulate that the application for a renewed license should ude a list of plant-specific exemptions granted pursuant to 10 CFR 50.12 and that are based on e-limited aging analyses, as defined in 10 CFR 54.3. Each active 10 CFR 50.12 exemption has n reviewed to determine whether the exemption is based on a time-limited aging analysis. No t-specific exemptions granted pursuant to 10 CFR 50.12 and based on a time-limited aging lyses as defined in 10 CFR 54.3 have been identified.

28/18 19.5-1 Rev. 31

le 19.6-1, License Renewal Commitments, provides a listing of the license renewal mitments.

.1 REFERENCES FOR SECTION 19.6

-37 Letter from Leslie N. Hartz to NRC, Millstone Power Station Units 2 and 3, Response to Request for Additional Information License Renewal Applications, August 13, 2004 (Serial No.: 04-398).

28/18 19.6-1 Rev. 31

Item Commitment Source Schedule a 1 The existing inspection program will be modified to include those battery racks Battery Rack Inspections Prior to Period of Extended Operation that require monitoring for license renewal, but are not already included in the program.

2 Implementing procedures will be modified to include loss of material as a Battery Rack Inspections Prior to Period of Extended Operation potential aging effect and to provide guidance on the inspection of items (such as anchorages, bracing and supports, side and end rails, and spacers), which contribute to battery rack integrity or seismic design of the battery racks.

3 A baseline inspection of the in-scope buried piping located in a damp soil Buried Pipe Inspection Program Prior to Period of Extended Operation environment will be performed for a representative sample of each combination of material and protective measures. Inspection for the loss of material due to selective leaching will be performed by visual, and mechanical or other appropriate methods.

4 The maintenance and work control procedures will be revised to ensure that Buried Pipe Inspection Program Prior to Period of Extended Operation inspections of buried piping are performed when the piping is excavated during maintenance or for any other reason. These procedures will include the inspection for the loss of material due to selective leaching which will be performed by visual, and mechanical or other appropriate methods.

5 The Electrical Cables and Connectors Not Subject to 10 CFR 50.49 Electrical Cables and Connectors Not Prior to Period of Extended Operation Environmental Qualification Requirements program will be established. Subject to 10 CFR 50.49 Environmental Qualification Requirements 6 Fuse holders meeting the requirements will be evaluated prior to the period of Electrical Cables and Connectors Not Complete.

extended operation for possible aging effects requiring management. The fuse Subject to 10 CFR 50.49 Environmental holder will either be replaced, modified to minimize the aging effects, or this Qualification Requirements program will manage the aging effects. The program (if needed for fuse holders) will consider the aging stressors for the metallic clips.

06/28/18 19.6-2 Rev. 31

Item Commitment Source Schedule a 7 Procedures will be developed to employ an alternate testing methodology to Electrical Cables Not Subject to Prior to Period of Extended Operation confirm the condition of cables and connectors in circuits that have sensitive, 10 CFR 50.49 Environmental Qualification Not to Exceed a 10 year Frequency low level signals and where the instrumentation is not calibrated in situ. Requirements Used in Instrumentation Thereafter Circuits 8 A baseline visual inspection will be performed on a representative sample of Fire Protection Program Prior to Period of Extended Operation the buried fire protection piping and components, whose internal surfaces are exposed to raw water, to confirm there is no degradation.

9 Testing a representative sample of fire protection sprinkler heads or replacing Fire Protection Program Prior to The Sprinkler Heads Achieving those that have been in service for 50 years will be included in the Fire 50 Years of Service Life Protection Program. Not to Exceed a 10 Year Interval Thereafter 10 The procedures and training for personnel performing General Condition General Condition Monitoring Prior to Period of Extended Operation Monitoring inspections and walkdowns will be enhanced to provide expectations that identify the requirements for the inspection of aging effects.

11 In-scope cable found to be submerged will be subject to an engineering Inaccessible Medium Voltage Cables Not Prior to Period of Extended Operation evaluation and corrective action. The evaluation of cables having significant Subject to 10 CFR 50.49 Environmental During the Corresponding 10 Year Interv voltage found to be submerged in standing water for an extended period of time Qualification Requirements (If Applicable) will be based on appropriate testing (using available technology consistent with NRC positions) of cables that are determined to be wetted for a significant period of time. The Engineering evaluation will also address the appropriate testing requirements for the corresponding ten-year intervals during the period of extended operation. The test will use a proven methodology for detecting deterioration of the insulation system due to wetting. Examples of such tests include power factor, partial discharge, or polarization index, as described in EPRI TR-103834-P1-2, Effects of Moisture on the Life of Power Plant Cables, or other appropriate testing. Testing will have acceptance criteria defined in accordance with the specific test identified. Occurrence of degradation that is adverse to quality is entered into the Corrective Action Program.

06/28/18 19.6-3 Rev. 31

Item Commitment Source Schedule a 12 The Infrequently Accessed Areas Inspection Program will be established. Infrequently Accessed Areas Inspection Prior to Period of Extended Operation Program 13 Millstone will follow the industry efforts on reactor vessel internals regarding Inservice Inspection Program: Reactor At Least Two Years Prior to Period of such issues as thermal or neutron irradiation embrittlement (loss of fracture Vessel Internals Extended Operation toughness), void swelling (change in dimensions), stress corrosion cracking (PWSCC and IASCC), and loss of pre-load for baffle and former-assembly bolts and will implement the appropriate recommendations resulting from this guidance. The revised program description, including a comparison to the 10 program elements of the NUREG-1801 program, will be submitted to the NRC for approval.

14 Augmented inspection of the Millstone Unit 3 core barrel holddown spring will Inservice Inspection Program: Reactor At Least Two Years Prior to the Period o be performed. In particular, the inspection will detect gross indication of loss of Vessel Internals Extended Operation preload as an aging effect. As an alternative to performing an augmented inspection, the holddown spring will be replaced.

15 Millstone will follow the industry efforts investigating the aging effects Inservice Inspection Program: Systems, At Least Two Years Prior to Period of applicable to nickel-based alloys (i.e., PWSCC in Alloy 600 base metal and Components and Supports Extended Operation Alloy 82/182 weld metals) and identifying the appropriate aging management Steam Generator Structural Integrity activities and will implement the appropriate recommendations resulting from this guidance. The revised program description will be submitted at least two years prior to the period of extended operation for staff review and approval to determine if the program demonstrates the ability to manage the effects of aging in nickel based components per 10 CFR 50.54.21(a)(3).

16 The existing inspection program will be modified to include those lifting Inspection Activities: Load Handling Cranes Prior to Period of Extended Operation devices that require monitoring for license renewal, but are not already and Devices included in the program.

17 Implementing procedures and documentation will be modified to include Inspection Activities: Load Handling Cranes Prior to Period of Extended Operation visual inspections for the loss of material on the crane and trolley structural and Devices components and the rails in the scope of license renewal in Commitment 16.

06/28/18 19.6-4 Rev. 31

Item Commitment Source Schedule a 18 The implementing procedures will be modified to include Structures Monitoring Program Prior to Period of Extended Operation ACI 349.3R-96 and ANSI/ASCE 11-90 as references and as input documents for the inspection program.

19 The Structures Monitoring Program and implementing procedures will be Structures Monitoring Program Prior to Period of Extended Operation modified to include all in-scope structures.

20 Groundwater samples will be taken on a periodic basis, considering seasonal Structures Monitoring Program Prior to Period of Extended Operation variations, to ensure that the groundwater is not sufficiently aggressive to cause the below-grade concrete to degrade.

21 The Structures Monitoring Program and implementing procedures will be Structures Monitoring Program Prior to Period of Extended Operation modified to alert the appropriate engineering organization if the structures inspections identify that medium voltage cables in the scope of license renewal have been submerged.

22 The maintenance and work control procedures will be revised to ensure that Structures Monitoring Program Prior to Period of Extended Operation inspections of inaccessible areas are performed when the areas become accessible by such means as excavation or installation of shielding during maintenance or for any other reason.

23 Appropriate inspections of sealants and caulkings used for moisture intrusion Tank Inspection Program Prior to Period of Extended Operation prevention in and around aboveground tanks will be performed.

24 Non-destructive volumetric examination of the in-scope inaccessible locations, Tank Inspection Program Prior to Period of Extended Operation such as the external surfaces of tank bottoms, will be performed prior to the A frequency consistent with scheduled period of extended operation. Subsequent inspections will be performed on a tank internals inspection activities frequency consistent with scheduled tank internals inspection activities.

25 The security diesel fuel oil tank and diesel fire pump fuel oil tank are in-scope Tank Inspection Program Complete for license renewal and will be included on the respective Tank Inspection Program inspection plan.

06/28/18 19.6-5 Rev. 31

Item Commitment Source Schedule a 26 Changes will be made to maintenance and work control procedures to ensure Work Control Process Prior to Period of Extended Operation that inspections of plant components and plant commodities will be appropriately and consistently performed and documented for aging effects during maintenance activities.

27 Consistent with 10 CFR 54.21(c)(1),(iii), the effects of environmentally Environmentally Assisted Fatigue TLAA Prior to Period of Extended Operation assisted fatigue for those specific locations with a CUF greater than 1.0 will be managed by the Metal Fatigue of Reactor Coolant Pressure Boundary program.

If the specific locations are not repaired, replaced, or successfully re-analyzed, a modified inspection program description, including a comparison to the 10 program elements of NUREG-1801 program, will submitted to the NRC for approval.

28 For potentially susceptible CASS materials, either enhanced volumetric Inservice Inspection Program: Systems, Prior to Period of Extended Operation examinations or a unit or component specific flaw tolerance evaluation Components and Supports (considering reduced fracture toughness and unit specific geometry and stress information) will be used to demonstrate that the thermally-embrittled material has adequate fracture toughness in accordance with NUREG-1801 Section XI.M12.3.

29 Millstone will follow industry efforts that will provide specific guidance to Environmentally Assisted Fatigue TLAA Prior to Period of Extended Operation license renewal applicants for evaluating the environmental effects of fatigue on applicable locations, other than those identified in NUREG/CR-6260.

Millstone will also implement the appropriate recommendations resulting from this guidance. Until these recommendations are available, Millstone 3 commits to using the pressurizer surge line nozzle as a leading indicator to address environmental effects on fatigue of pressurizer sub-components during the period of extended operation.

06/28/18 19.6-6 Rev. 31

Item Commitment Source Schedule a 30 A baseline visual inspection will be performed of the accessible areas of the Closed-Cycle Cooling Water System Prior to Period of Extended Operation shell side (including accessible portions of the exterior side of the tubes) of one:

Millstone Unit 2 Reactor Building Closed Cooling Water heat exchanger, Millstone Unit 2 Emergency Diesel Generator Jacket Cooling Water heat exchanger, and Millstone Unit 3 Emergency Diesel Generator Jacket Cooling Water heat exchanger.

31 Using the Work Control Process, a baseline inspection for the loss of material Work Control Process Prior to Period of Extended Operation due to selective leaching will be performed on a representative sample of locations for susceptible materials by visual, and mechanical or other appropriate methods prior to entering the period of extended operation.

32 A review of the Work Control Process inspection opportunities for each Work Control Process Prior to Period of Extended Operation material and environment group, supplemental to the initial review conducted during the development of the LRA, will be performed. Baseline inspections will be performed for the material and environment combinations that have not been inspected as part of the Work Control Process.

33 Calibration results for cables tested in situ will be reviewed to detect severe Electrical Cables Not Subject to Prior to Period of Extended Operation aging degradation of the cable insulation. The initial review will be completed 10 CFR 50.49 Environmental Qualification Not to Exceed a 10 Year Frequency prior to entering the period of extended operation and will include at least 5 Requirements Used in Instrumentation Thereafter years of surveillance test data for each cable reviewed. Subsequent reviews Circuits will be performed on a period not to exceed 10 years.

34 The in scope cables in Unit 3 duct lines # 929 (SBO Diesel to Unit 3 4.16kV Inaccessible Medium Voltage Cables Not Prior to period of extended operation.

Normal Switchgear) and # 973 (RSST 3RTX-XSR-B to 6.9kV Normal Subject to 10 CFR 50.49 Environmental Complete. Subsequent testing will not Switchgear Bus 35A, 35B, 35C and 35D) will be tested to demonstrate that Qualification Requirements exceed a 10 year frequency.

water treeing will not prevent the cables from performing their intended function.

06/28/18 19.6-7 Rev. 31

Item Commitment Source Schedule a 35 In addition to the testing specified in Commitment 34, a representative sample Inaccessible Medium Voltage Cables Not Prior to Period of Extended Operation of in-scope medium voltage cables will be tested to demonstrate that water Subject to 10 CFR 50.49 Environmental Not to Exceed a 10 Year Frequency treeing will not prevent the cables from performing their intended function. Qualification Requirements Thereafter 36 Millstone Unit 3 will complete the SAMA evaluation of the ability to manually Severe Accident Mitigation Alternatives Prior to Period of Extended Operation control the Turbine Driven Auxiliary Feedwater Pump. If this SAMA is cost (SAMA) Analysis (Reference 19.6-37) beneficial (i.e., can be accomplished without a hardware modification), a Severe Accident Management Guideline (SAMG) addressing this mitigation strategy will be developed.

37 The pressurizer spray head assembly will be either replaced or inspected Inservice Inspection Program: Systems, Prior to Period of Extended Operation utilizing the best currently available (at the time of inspection) techniques for Components and Supports detecting cracking resulting from SCC.

a. The period of extended operation is the period of 20 years beyond the expiration date of the Units previous 40 year Operating License.

06/28/18 19.6-8 Rev. 31

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