ML17212A058

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Final Safety Analysis Report, Rev. 35, Chapter 15, License Renewal
ML17212A058
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Site: Millstone Dominion icon.png
Issue date: 06/29/2017
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Dominion Nuclear Connecticut
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Office of Nuclear Reactor Regulation
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Millstone Power Station Unit 2 Safety Analysis Report Chapter 15

Table of Contents tion Title Page Introduction........................................................................................................ 15.0-1 Overview of Aging Management Programs ...................................................... 15.1-1

.1 Aging Management Programs .................................................................. 15.1-1

.2 Time Limited Aging Analyses Aging Management Programs: ............... 15.1-2

.3 Aging Management Programs Created After Issuance of the Renewed Operating License ..................................................................................... 15.1-2

.4 References for Section 15.1 ...................................................................... 15.1-3 Programs that Manage the Effects of Aging on Structures and Components Within the Scope of License Renewal ................................................................................. 15.2-1

.1 Aging Management Programs .................................................................. 15.2-1

.1.1 Battery Rack Inspections .......................................................................... 15.2-1

.1.2 Boraflex Monitoring ................................................................................. 15.2-2

.1.3 Boric Acid Corrosion................................................................................ 15.2-2

.1.4 Buried Pipe Inspection Program ............................................................... 15.2-3 2.1.5 Chemistry Control for Primary Systems Program .................................... 15.2-4

.1.6 Chemistry Control for Secondary Systems Program ................................ 15.2-4

.1.7 Closed-Cycle Cooling Water System ....................................................... 15.2-4

.1.8 Electrical Cables and Connectors Not Subject to 10 CFR 50.49 Environmental Qualification Requirements ...................................................................... 15.2-5

.1.9 Electrical Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits ....................................... 15.2-6

.1.10 Fire Protection Program............................................................................ 15.2-7

.1.11 Flow-Accelerated Corrosion..................................................................... 15.2-8

.1.12 Fuel Oil Chemistry.................................................................................... 15.2-9

.1.13 General Condition Monitoring.................................................................. 15.2-9

.1.14 Inaccessible Medium Voltage Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements ........................................... 15.2-10 2.1.15 Infrequently Accessed Areas Inspection Program.................................. 15.2-11

.1.16 Inservice Inspection Program: Containment Inspections ....................... 15.2-12

.1.17 Inservice Inspection Program: Reactor Vessel Internals ........................ 15.2-13

.1.18 Inservice Inspection Program: Systems, Components and Supports...... 15.2-14

.1.19 Inspection Activities: Load Handling Cranes and Devices .................... 15.2-15 2.1.20 Reactor Vessel Surveillance ................................................................... 15.2-16

.1.21 Service Water System (Open-Cycle Cooling) ........................................ 15.2-16

.1.22 Steam Generator Structural Integrity ...................................................... 15.2-17

.1.23 Structures Monitoring Program .............................................................. 15.2-17 15-i Rev. 35

tion Title Page

.1.24 Tank Inspection Program........................................................................ 15.2-19

.1.25 Work Control Process ............................................................................. 15.2-20

.1.26 Bolting Integrity Program ....................................................................... 15.2-21

.1.27 Alloy 600 Management Program............................................................ 15.2-22

.2 References for Section 15.2 .................................................................... 15.2-22 Time-Limited Aging Analysis ........................................................................... 15.3-1 3.1 Reactor Vessel Neutron Embrittlement .................................................... 15.3-1

.1.1 Upper Shelf Energy .................................................................................. 15.3-1

.1.2 Pressurized Thermal Shock ...................................................................... 15.3-1

.1.3 Pressure-Temperature Limits.................................................................... 15.3-2

.2 Metal Fatigue ............................................................................................ 15.3-3

.2.1 Millstone Unit 2 Class 1 Components ...................................................... 15.3-3

.2.2 Non-Class 1 Components ......................................................................... 15.3-4

.2.3 Environmentally Assisted Fatigue ............................................................ 15.3-4

.3 Environmental Qualification (EQ) of Electric Equipment ....................... 15.3-5

.4 Concrete Containment Tendon Prestress .................................................. 15.3-6

.5 Containment Liner Plate, Metal Containments, and Penetrations Fatigue Analysis .................................................................................................... 15.3-6

.5.1 Containment Liner Plate ........................................................................... 15.3-6 3.5.2 Containment Penetrations ......................................................................... 15.3-7

.6 Other Plant-Specific Time-Limited Aging Analyses................................ 15.3-7

.6.1 Crane Load Cycle Limit ........................................................................... 15.3-7

.6.2 Reactor Coolant Pump Flywheel .............................................................. 15.3-7

.6.3 Reactor Coolant Pump Code Case N-481................................................. 15.3-8

.6.4 Leak-Before-Break ................................................................................... 15.3-8

.7 References for Section 15.3 ...................................................................... 15.3-9 TLAA Support programs ................................................................................... 15.4-1

.1 Electrical Equipment Qualification .......................................................... 15.4-1

.2 Metal Fatigue of Reactor Coolant Pressure Boundary ............................. 15.4-1 Exemptions ........................................................................................................ 15.5-1 6 License Renewal Commitments ........................................................................ 15.6-1

.1 References for Section 15.6 ...................................................................... 15.6-1 15-ii Rev. 35

List of Tables mber Title

-1 Containment Tendon Prestress

-1 License Renewal Commitments 15-iii Rev. 35

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 15.1 contains a listing of the aging management programs and the status of the program at the time the License Renewal Application was submitted. This section also contains a list of new aging management programs created after the issuance of the renewed operating license.

Section 15.2 contains a description of the programs for managing the effects of aging.

Section 15.3 contains the evaluation of Time-limited Aging Analyses (TLAAs) for the period of extended operation.

Section 15.4 contains a summarized description of the programs that support the TLAAs.

Section 15.5 contains a summarized description of the plant-specific exemptions.

Section 15.6 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.

15.0-1 Rev. 35

1.1 AGING MANAGEMENT PROGRAMS aging management programs for Millstone Unit 2 are described in the following sections. The grams are either consistent with generally accepted industry methods as discussed in REG-1801 (Reference 15.1-1), 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 new 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 15.2.1.1) (Existing - Requires Enhancement).
2. Boraflex Monitoring (Section 15.2.1.2) (Existing).
3. Boric Acid Corrosion (Section 15.2.1.3) (Existing).
4. Buried Pipe Inspection Program (Section 15.2.1.4) (Existing - Requires Enhancement).
5. Chemistry Control for Primary Systems Program (Section 15.2.1.5) (Existing).
6. Chemistry Control for Secondary Systems Program (Section 15.2.1.6) (Existing).
7. Closed-Cycle Cooling Water System (Section 15.2.1.7) (Existing - Requires Enhancement).
8. Electrical Cables and Connectors Not Subject to 10 CFR 50.49 (Section 15.2.1.8)

(To Be Developed).

9. Electrical Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements used in Instrumentation Circuits (Section 15.2.1.9) (Existing -

Requires Enhancement).

10. Fire Protection Program (Section 15.2.1.10) (Existing - Requires Enhancement).
11. Flow-Accelerated Corrosion (Section 15.2.1.11) (Existing).
12. Fuel Oil Chemistry (Section 15.2.1.12) (Existing).
13. General Condition Monitoring (Section 15.2.1.13) (Existing - Requires Enhancement).

15.1-1 Rev. 35

Enhancement).

15. Infrequently Accessed Areas Inspection Program (Section 15.2.1.15) (To Be Developed).
16. Inservice Inspection Program: Containment Inspections (Section 15.2.1.16)

(Existing).

17. Inservice Inspection Program: Reactor Vessel Internals (Section 15.2.1.17)

(Existing - Requires Enhancement).

18. Inservice Inspection Program: Systems, Components and Supports (Section 15.2.1.18) (Existing - Enhancement complete).
19. Inspection Activities: Load Handling Cranes and Devices (Section 15.2.1.19)

(Existing - Requires Enhancement).

20. Reactor Vessel Surveillance (Section 15.2.1.20) (Existing).
21. Service Water System (Open-Cycle Cooling) (Section 15.2.1.21) (Existing).
22. Steam Generator Structural Integrity (Section 15.2.1.22) (Existing).
23. Structures Monitoring Program (Section 15.2.1.23) (Existing - Requires Enhancement).
24. Tank Inspection Program (Section 15.2.1.24) (Existing - Requires Enhancement).
25. Work Control Process (Section 15.2.1.25) (Existing - Requires Enhancement).
26. Bolting Integrity Program (Section 15.2.1.26) (Existing).

1.2 TIME LIMITED AGING ANALYSES AGING MANAGEMENT PROGRAMS:

1. Electrical Equipment Qualification (Section 15.4.1) (Existing).
2. Metal Fatigue of Reactor Coolant Pressure Boundary (Section 15.4.2) (Existing).

.3 AGING MANAGEMENT PROGRAMS CREATED AFTER ISSUANCE OF THE RENEWED OPERATING LICENSE

1. Alloy 600 Management Program (Section 15.2.1.27).

15.1-2 Rev. 35

-1 NUREG-1801, Generic Aging Lessons Learned (GALL) Report, U. S. Nuclear Regulatory Commission, July 2001 15.1-3 Rev. 35

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 has been modified to include those battery racks that require monitoring for license renewal, but were not originally included in the program.

Inclusion of those battery racks completes the actions required for Commitment Item 1 in Table 15.6-1.

Inspection Criteria Implementing procedures have been 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 15.2-1 Rev. 35

.1.2 Boraflex Monitoring gram Description aflex is no longer credited as a neutron absorber. Thus, Boraflex monitoring is no longer ormed.

.1.3 Boric Acid Corrosion 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 mary 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 15.2-15)

NRC Bulletin 2002-01 (Reference 15.2-16)

NRC Bulletin 2002-02 (Reference 15.2-17)

NRC Bulletin 2003-02 (Reference 15.2-18)

NRC Order EA-03-009 (Reference 15.2-19)

NRC Bulletin 2004-01 (Reference 15.2-20) 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.

15.2-2 Rev. 35

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.

re are no buried tanks within the scope of license renewal.

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 has been performed for a representative sample of each combination of material and protective measures. Inspection for the loss of material due to selective leaching has been performed by visual, and mechanical or other appropriate methods. These inspections complete the actions required for Commitment Item 3 in Table 15.6-1.

Buried Piping Inspections The maintenance and work control procedures have been revised to ensure that inspections of buried piping are performed when the piping is excavated during maintenance or for any other reason. These procedures include the inspection for the loss of material due to selective leaching, which will be performed by visual, and mechanical or other appropriate methods. These changes complete the actions required for Commitment Item 4 in Table 15.6-1.

15.2-3 Rev. 35

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 15.2-1) 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 urance Program. The corrective action process provides reasonable assurance that deficiencies erse to quality are either promptly corrected or are evaluated to be acceptable.

2.1.6 Chemistry Control for Secondary Systems Program gram Description mistry Control for Secondary Systems Program corresponds to NUREG-1801, tion XI.M2, Water Chemistry. The program includes periodic monitoring and control of wn detrimental contaminants such as chlorides, sodium, dissolved oxygen, and sulfate centrations below the levels known to result in loss of material or cracking. Water chemistry trol is in accordance with the guidelines in EPRI TR-102134 (Reference 15.2-2) for secondary er 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 urance Program. The corrective action process provides reasonable assurance that deficiencies erse to quality are either promptly corrected or are evaluated to be acceptable.

2.1.7 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 15.2-3).

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 15.2-4 Rev. 35

mmitments following commitment has been implemented prior to the period of extended operation:

Heat Exchanger Baseline Inspection A baseline visual inspection has been 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,

performance of these inspections complete the actions required for Commitment Item 29 in le 15.6-1.

2.1.8 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 15.2-4). 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.

15.2-5 Rev. 35

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 has been established.

This completes the action required to complete Commitment Item 5 in Table 15.6-1, License Renewal Commitments.

Inclusion of In-Scope Fuse Holders Fuse holders meeting the requirements have been evaluated. The fuse holder will either be replaced, modified to minimize the aging effects, or this program will manage the aging effects. The aging management review of fuse holders considered the aging stressors for metallic clips and concluded that there were no aging effects that require management.

This completes the actions required to complete Commitment Item 6 in Table 15.6-1, License Renewal Commitments.

.1.9 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 15.2-5). 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.

15.2-6 Rev. 35

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 have been 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 have been completed.

The frequency of subsequent tests will be based on Engineering evaluation and will not exceed a 10 year interval. This completes the action required to complete commitment Item 7 in Table 15.6-1, License Renewal Commitments.

Review of Surveillance Test Results for Cables Tested In Situ Calibration results for cables tested in situ have been reviewed to detect severe aging degradation of the cable insulation, and include at least 5 years of surveillance test data for each cable reviewed. Subsequent reviews will be performed on a period not to exceed 10 years. This completes the action required to complete commitment Item 32 in Table 15.6-1, License Renewal Commitments.

.1.10 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 15.2-6).

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.

15.2-7 Rev. 35

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 A baseline visual inspection has been performed on a representative sample of the buried fire protection piping and components, whose internal surfaces are exposed to raw water, to confirm there is no degradation. These inspections complete the action required for Commitment Item 8 in Table 15.6-1.

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 has been 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 change completes the action required for Commitment Item 9 in Table 15.6-1, License Renewal Commitments.

.1.11 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 15.2-7). 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.

15.2-8 Rev. 35

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.

.1.12 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:

TM Standard D 1796 (Reference 15.2-8),

TM Standard D 2276 (unmodified) (Reference 15.2-9), and TM Standard D 4057 (Reference 15.2-10).

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.

.1.13 General Condition Monitoring gram Description eral Condition Monitoring is a plant specific program that manages the aging effects of loss of erial, change of material properties, and cracking on the external surfaces of components. It is ormed 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 Health Physics technicians, System Engineers, 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 15.2-9 Rev. 35

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 have been enhanced to provide expectations that identify the requirements for the inspection of aging effects. These actions complete the requirements for Item 10 in Table 15.6-1.

.1.14 Inaccessible Medium Voltage Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements 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 o 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:

15.2-10 Rev. 35

In scope cable found to be submerged has been 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 was 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 also addresses the appropriate testing requirements for the corresponding ten-year intervals during the period of extended operation. The test used 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 Cables, or other appropriate testing.

Testing has acceptance criteria defined in accordance with the specific test identified.

Occurrence of degradation that is adverse to quality was entered into the Corrective Actions Program. This completes the actions required to complete commitment Item 11 in Table 15.6-1, License Renewal Commitments.

Testing of Inaccessible Medium Voltage Cables The in-scope cables in Unit 3 duct lines number 929 (SBO Diesel to Unit 3 4.16kV Normal Switchgear) and number 973 (RSST 3RTX-XSR-B to 6.9kV Normal Switchgear Bus 35A, 35B, 35C, 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 33 in Table 15.6-1, License Renewal Commitments.

Sample Testing of Inaccessible Medium Voltage Cables In addition to the testing specified in Commitment 33, a representative sample of in-scope medium voltage cables has 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.

.1.15 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 15.2-11 Rev. 35

er promptly corrected or are evaluated to be acceptable.

mmitments following program enhancements have been implemented prior to the period of extended ration:

Program Implementation The Infrequently Accessed Inspection Program has been established.

The establishment of the Infrequently Accessed Inspection Program completes the action required for Commitment Item 12 in Table 15.6-1.

.1.16 Inservice Inspection Program: Containment Inspections gram Description rvice Inspection Program: Containment Inspections corresponds to the following REG-1801 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.

15.2-12 Rev. 35

gram Description rvice Inspection Program: Reactor Vessel Internals corresponds to the following REG-1801 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.

Inservice Inspection Program: Reactor Vessel Internals manages the effects of aging for those tor internals that are susceptible to loss of material, cracking, loss of preload, change in 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 preload 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 at 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), and stress corrosion cracking (PWSCC and IASCC) 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, has been submitted to the NRC for approval.

The submission of the program description to the NRC completes the required actions for Commitment Item 13.

15.2-13 Rev. 35

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.M12, Thermal Aging Embrittlement of Cast Austentic 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 15.2-11). 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.

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:

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.

15.2-14 Rev. 35

that the surge line is highly flaw tolerant but recommended volumetric examinations to confirm that no flaws exist that are deeper than the allowable flaw depths. Successful inspections of the pressurizer surge line during the M2R22 refueling outage confirmed that no adverse flaws exist. Those inspections complete the actions required for Commitment Item 27, Table 15.6-1.

Pressurizer Replacement Dominion has replaced the Millstone Unit 2 pressurizer using materials that are resistant to PWSCC. Replacing the pressurizer completes the action required for Commitment Item 36.

.1.19 Inspection Activities: Load Handling Cranes and Devices gram Description ection Activities: Load Handling Cranes and Devices corresponds to NUREG-1801, tion XI. M23, Inspection of Overhead Heavy Load [Related to Refueling] Handling tems. The program manages the aging effect of loss of material for the load handling cranes devices within the scope of license renewal. The in-scope load handling cranes and devices either safety-related or seismically designed to ensure that they will not adversely impact ty-related components 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 base plates 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.

mmitments following program enhancements have been implemented prior to the period of extended ration:

Inclusion of In-Scope Lifting Devices The existing inspection program has been modified to include all lifting devices that require monitoring for license renewal. This commitment is identified in Table 15.6-1, License Renewal Commitments, Item 15.

15.2-15 Rev. 35

Implementing procedures and documentation have been 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 15.6-1, License Renewal Commitments, Item 16.

.1.20 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.

.1.21 Service Water System (Open-Cycle Cooling) gram Description Service Water System (Open-Cycle Cooling) program corresponds to NUREG-1801, tion XI.M20, Open Cycle Cooling Water System. The program manages the aging effects of of material and buildup of deposits. The program implements the NRC guidelines in Generic er 89-13 (Reference 15.2-12), which includes (a) surveillance and control of biofouling; (b) a program to verify heat transfer capabilities; (c) routine inspection and a maintenance program 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. In lieu of thermal performance testing, Millstone Unit 2 relies on frequent, regular ection and cleaning of heat exchangers 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 15.2-16 Rev. 35

er promptly corrected or are evaluated to be acceptable.

.1.22 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 15.2-13). 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.

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.

.1.23 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 15.2-14). 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.

15.2-17 Rev. 35

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:

Modification of Structures Monitoring Program procedures 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 procedure for the Structures Monitoring Program has been modified to include ACI 349.3R-96 and ANSI/ASCE 11-90 as references and as input documents for the inspection program. Modification of the implementing procedure completes the action required for Commitment Item 17 in Table 15.6-1.

Addition of Structures to the Structures Monitoring Program The Structures Monitoring Program did not initially monitor all structures in-scope for license renewal. The Structures Monitoring Program and the implementing procedure have been modified to include all in-scope structures. Modification of the Program and the implementing procedure completes the actions required for Commitment Item 18 in Table 15.6-1.

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. The implementing procedure for the Structures Monitoring Program has been revised to direct groundwater sampling. Modification of the implementing procedure completes the action required for Commitment Item 19 in Table 15.6-1.

Engineering Notification of Submerged Medium Voltage Cables The Structures Monitoring Program and implementing procedures have been 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.

15.2-18 Rev. 35

Inspection of Normally Inaccessible Areas That Become Accessible Initially, the license renewal implementation plans stated that 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. The procedure revision task is accomplished by a modification of the Structures Monitoring implementing procedure for the situation of an inaccessible area becoming accessible by excavation or installation of shielding. Modification of the implementing procedure completes the action required for Commitment Item 21 in Table 15.6-1.

.1.24 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, ngs, and structural elements, to prevent deterioration of the tanks to unacceptable levels. The gram also includes volumetric examination of inaccessible locations, such as the external 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 have been performed. These inspections complete the actions required for Commitment Item 22 in Table 15.6-1.

Non-destructive Volumetric Examination of Inaccessible Tank Bottoms 15.2-19 Rev. 35

operation. Subsequent inspections will be performed on a frequency consistent with scheduled tank internals inspection activities. These examinations and scheduled subsequent inspections complete the actions required for Commitment Item 23 in Table 15.6-1.

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 actions required for Commitment Item 24 in Table 15.6-1.

.1.25 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 t commodities within the scope of license renewal. Performance testing and maintenance vities, both preventive and corrective, are planned and conducted in accordance with the Work 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 15.2-20 Rev. 35

consistently performed and documented for aging effects during maintenance activities.

These changes complete the actions required for Commitment Item 25 in Table 15.6-1.

Selective Leaching Inspection Using the Work Control Process, a baseline inspection for the loss of material due to selective leaching have been performed on a representative sample of buried pipe locations for susceptible materials by visual, and mechanical or other appropriate methods prior to entering the period of extended operation. Susceptible aluminum-bronze valves have been replaced using a material (AL6XN stainless steel) that is more resistant to selective leaching. The buried pipe inspections and valve replacements complete the actions required for Commitment Item 30 in Table 15.6-1.

Verification of Scope A review of the Work Control Process inspection opportunities for each material and environment group supplemental to the initial review conducted during the development of the LRA has been performed. No additional baseline inspections were needed to be performed for the material and environment combinations inspected as part of the Work Control Process. This action completes the requirements for Item 31 in Table 15.6-1.

.1.26 Bolting Integrity Program 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 TR-13, Bolted Joint Maintenance and Application Guide. For ASME Class bolting, aging effects additionally managed by the performance of inservice examinations in accordance with ME 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.

15.2-21 Rev. 35

gram Description Alloy 600 Management Program manages the aging effect of primary water stress corrosion king (PWSCC) in Alloy 600 base metal and Alloy 82/182 dissimilar metal welds. The Alloy Management Program is consistent with the Nickel-Alloy Nozzles and Penetrations aging agement program described in Chapter XI of NUREG-1801, Rev. 0 (GALL Report). The gram monitors susceptible components for indications of PWSCC before there is a loss of nded function. Since the program is based on more recent industry guidance, EPRI TR-9561, MRP-126 (Reference 15.2-21), the program is enhanced from the program described in

. 0 of the GALL. The program scope has been increased according to industry guidelines to ude all Alloy 600/82/182 components in the primary system. The inspection requirements e been updated to the requirements of ASME Section XI (Reference 15.2-11) Code Cases 22-1, N-729-1 and N-770-1.

Alloy 600 Management Program credits inspections performed under existing inspection cesses. Inspection requirements for each type of component are listed in the Alloy 600 nagement Plan. Defects found during inspections will be dispositioned in accordance with licable regulatory and code requirements, utilizing the corrective action system.

mmitments following action will be implemented at least two years prior to the period of extended ration:

PWSCC of Nickel Based Alloys Millstone will follow the industry efforts investigating the aging effects applicable to nickel based alloys (i.e., PWSCC in Alloy 600 base metal and Alloy 82/182 weld metals) and identifying the appropriate aging management activities and will implement the appropriate recommendations resulting from this guidance. The revised program description was submitted 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 54.21(a)(3).

The submission of the Alloy 600 Management Program description to the NRC (Reference 15.6-3) completes the required actions for Commitment Item 14 in Table 15.6-1.

2.2 REFERENCES

FOR SECTION 15.2

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

15.2-22 Rev. 35

-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.

2-6 NRC Interim Staff Guidance (ISG)-04, Aging Management of Fire Protection Systems for License Renewal, U.S. Nuclear Regulatory Commission, December 3, 2002.

-7 NSAC-202L-R4 Recommendations for an Effective Flow Accelerated Corrosion Program, Electric Power Research Institute, November, 2013.

-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.

15.2-23 Rev. 35

-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.

-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.

-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.

-21 EPRI Report 1009561, Materials Reliability Program: Generic Guidance for Alloy 600 Management (MRP-126), Electric Power Research Institute, Palo Alto, CA November 2004.

15.2-24 Rev. 35

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 4.3.1. Time-limited aging analyses (TLAAs) licable to the reactor vessel are:

Upper-shelf energy (USE)

Pressurized thermal shock (PTS)

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 f energy 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 ft-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 an 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 de 1.99, Revision 2 to the end of the period of extended operation. Calculated upper shelf rgy values for the most limiting reactor pressure vessel beltline plate and weld materials ain greater than 50 ft-lbs.

3.1.2 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 15.3-1 Rev. 35

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 reference temperature ductility transition or adjusted reference temperature (adjusted RTNDT or ART) corresponding he limiting beltline material of the reactor pressure vessel for the current period of licensed ration. Current low temperature overpressure protection (LTOP) system heatup and cooldown t curves were approved in License Amendment 218.

ccordance with 10 CFR 50, Appendix G, updated pressure-temperature limits for entering the od of extended operation have been developed and implemented prior to the period of nded operation. Low temperature overpressure protection system enable temperature uirements were found to remain valid and will ensure that the pressure-temperature limits will be exceeded for postulated plant transients during the period of extended operation. Millstone t 2 will calculate USE, RTPTS, and P-T limits based on fluence values developed in ordance with Regulatory Guide 1.190 requirements, as amended or superseded by future ulatory guidance changes, through the period of extended operation.

mmitments following actions will be taken prior to the period of extended operation.

ated USE, RTPTS, and P-T limits based on fluence values developed in accordance with ulatory Guide 1.190 requirements, as amended or superseded by regulatory guidance changes, e required to be submitted to the NRC for review at least two years prior to the period of nded operation. This commitment is identified in Table 15.6-1, License Renewal mmitments, Item 37.

15.3-2 Rev. 35

ulatory Guide 1.190 requirements were submitted to the NRC for review in 2005 ference 15.6-4) and approved in 2006 (Reference 15.6-5).

submittal of the updated USE, RTPTS, and P-T limits completes the required actions for mmitment Item 37 in Table 15.6-1.

3.2 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 ber of cycles and the associated fatigue (cumulative) usage factors rather than specific time ts.

.2.1 Millstone Unit 2 Class 1 Components mponents within the Millstone Unit 2 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 15.3-1) and ANSI Standard B31.7 (Reference 15.3-2). Use of e codes requires that design analyses for Class A (Class 1) systems and components address gue and the establishment of load limits to preclude initiation of fatigue 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 2 response (Reference 15.3-3) and supplemental communications, NRC concluded that Millstone Unit 2 meets the requirements of Bulletin 88-08 ference 15.3-4).

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 Combustion Engineering ners Group task (Reference 15.3-5) supplemented by additional unit specific inspections and vities. Based upon the Combustion Engineering Owners Group task, the NRC concluded that owners group analysis CEN-387-P is bounding for thermal stratification and thermal striping 15.3-3 Rev. 35

C (Reference 15.3-6).

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 Austenetic 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.

eptable thermal and pressure transients, and operating cycles have been projected for ASME tion III Class A, Class 1 and ANSI Standard B31.7 Class 1 components though the period of nded operation.

3.2.2 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 15.3-8) 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 15.3-9). The evaluations associated with the newer-vintage mbustion Engineering plants are applicable, since the majority of the Millstone Unit 2 Class 1 ems and components were designed to ASME Section III/ANSI B31.7 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.

15.3-4 Rev. 35

Surge line.

Charging System nozzle.

Safety Injection System nozzle.

Shutdown Cooling line.

se six fatigue-sensitive locations have been evaluated using the methods identified in REG/CR-6583 (Reference 15.3-10), and NUREG/CR-5704 (Reference 15.3-11).

izing Millstone Unit 2 cyclic and transient information, all six fatigue sensitive component tions were determined to be acceptable for the period of extended operation.

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

Millstone follows industry efforts that 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 implement the appropriate recommendations resulting from this guidance. This action completes the requirements for Commitment Item 28 in Table 15.6-1.

3.3 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 is provided in tion 15.4.1, Electrical Equipment Qualification.

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 15.3-12). 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 15.3-13). Further guidance for post-accident monitoring equipment is 15.3-5 Rev. 35

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

3.4 CONCRETE CONTAINMENT TENDON PRESTRESS Millstone Unit 2 Containment consists of a pre-stressed, reinforced concrete cylinder and e, and a flat, reinforced concrete mat foundation supported on unweathered bedrock. The ndrical portion of the Containment is prestressed by a post-tensioning system composed of zontal and vertical tendons, with the horizontal tendons placed in three 240 degree systems use three buttresses as support for the anchorages. The dome has a three-way post tensioning em. Prestress on the containment tendons is expected to decrease over the life of the unit as a lt of such factors as elastic deformation, creep and shrinkage of concrete, anchorage seating es, tendon wire friction, stress relaxation and corrosion.

tainment tendon surveillance results has been performed and indicate acceptable losses in tainment tendon prestress have been projected through the period of extended operation.

ection results, Table 15.3-1, were used to project acceptable dome, vertical and horizontal es through the current and extended period of operation. Prestress values are projected to ain above the minimum value of 1308 kips (7.03 kips per wire) for each of the tendon groups.

ce containment tendon examinations are performed on a 5 year interval, force projections will efined and updated as necessary following each examination.

evaluation of containment tendon examination and surveillance test results involves the use ime-limited assumptions such as corrosion rates, losses of tendon prestress, and changes in erial properties. Regression analysis incorporating the most recent 25 years.

3.5 CONTAINMENT LINER PLATE, METAL CONTAINMENTS, AND PENETRATIONS FATIGUE ANALYSIS 3.5.1 Containment Liner Plate lstone Unit 2 has a prestressed, post-tensioned concrete Containment surrounded by an losure building. A welded carbon steel liner plate is attached to the inside surface of the crete. Both the liner plate and the penetration sleeves are designed to serve as the primary tainment leakage barrier. Components of the liner plate include penetration sleeves, access nings, piping penetrations, and electrical penetrations.

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.

15.3-6 Rev. 35

Millstone Unit 2 Containment penetrations are pressure resistant, leak-tight, welded mblies, fabricated, installed, inspected, and tested in accordance with ASME Nuclear Vessel e,Section III (Reference 15.3-1) and ANSI Nuclear Piping Code B31.7 (Reference 15.3-2).

luations of Containment liner plate components involve the use of time-limited assumptions h as corrosion rates and thermal cycles. These evaluations meet the requirements of CFR 54.3 and, as such, represent time-limited aging analyses.

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

3.6 OTHER PLANT-SPECIFIC TIME-LIMITED AGING ANALYSES 3.6.1 Crane Load Cycle Limit containment polar crane, spent fuel crane, and monorails are examples of the types of cranes hin the scope of license renewal. These cranes meet the guidance contained in NUREG-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 Number 70.

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

3.6.2 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.

evaluation was performed of the likelihood of flywheel failure over a 60 year period of ration and a justification was developed for relaxation of Regulatory Guide 1.14, Revision 1, ulatory Position C.4.b(1), requirements to those identified in Regulatory Position C.4.b(2) ference 15.3-15). Using this evaluation, the NRC issued Amendment Number 264 to the unit hnical Specifications, consistent with Regulatory Guide 1.14, Revision 1, Regulatory Position

.b(2), to allow the examination of each reactor coolant pump flywheel at least once every 10 rs, coinciding with the ASME Section XI inservice inspection program schedule.

15.3-7 Rev. 35

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, mponents and Supports, provides reasonable assurance that flywheel cracking will be quately managed for the period of extended operation.

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

3.6.3 Reactor Coolant Pump Code Case N-481 ME Boiler and Pressure Vessel Code,Section XI, specifies that a volumetric inspection of the tor coolant pump casing welds and a visual inspection of pump casing internal surfaces be ormed on a reactor coolant pump within each ten year inspection period. These 10 year umetric inspections are significant for a number of reasons, including; the reactor coolant ps are welded to the piping, and the pumps must be disassembled in order to gain access to inside surface of the cast austenitic stainless steel casings. In recognition of these difficulties, ME Code Case N-481, Alternative Examination Requirements for Cast Austenitic Pump ings, was developed to allow for the replacement of volumetric examinations with a fracture hanics-based evaluation, supplemented by specific visual inspections.

evaluation of reactor coolant pump casings 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 casing fatigue failure r a 60 year period, along with implementation of the Inservice Inspection Program: Systems, mponents and Supports, provides reasonable assurance that cracking of cast stainless steel tor coolant pump casing welds will be adequately managed throughout the period of extended ration.

3.6.4 Leak-Before-Break Leak-Before-Break (LBB) analyses were evaluated as time-limited aging analyses (TLAAs) etermine that the analyses remain valid for the period of extended operation. The systems and ponents that have been analyzed for LBB include the reactor coolant loop piping (hot leg, leg, and crossover piping), the pressurizer surge line, and portions of the safety injection and tdown cooling systems.

LBB analyses were 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 15.3.2.1 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.

15.3-8 Rev. 35

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 analyses 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 analyses have been projected to remain valid through the end of the period of extended ration.

3.7 REFERENCES

FOR SECTION 15.3

-1 ASME Section III, Rules for Construction of Nuclear Vessels, ASME Boiler and Vessel Pressure Code, American Society of Mechanical Engineers, 1971.

-2 ANSI Standard B31.7, Nuclear Power Piping Code, American Society of Mechanical Engineers.

-3 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.

-4 Letter from G.S. Vissing to E. J. Mroczka, NRC Bulletin 88-08, Thermal Stresses in Piping Connected to Reactor Coolant Systems (TAC 69652), September 30, 1991.

-5 Letter from E. J. Mroczka to NRC, NRC Bulletin Number 88-11, Pressurizer Surge Line Thermal Stratification, February 28, 1989.

-6 Letter from J. F. Opeka to NRC, Response to NRC Staff Request for Additional Information on Pressurizer Surge Line Thermal Stratification Evaluation, September 30, 1993.

-7 Letter from G. S. Vissing to J. F. Opeka, Safety Evaluation for Combustion Engineering Owners Group Report CEN-387-P, Revision 1 - Pressurizer Surge Line Thermal Stratification Evaluation (Bulletin 88-11)(TAC No. M72144), July 6, 1993.

-8 ANSI B31.1, Power Piping Code, American Society of Mechanical Engineers, 1967.

15.3-9 Rev. 35

-10 NUREG/CR-6583, Effects of LWR Coolant Environments on Fatigue Design Curves of Carbon and Low-Alloy Steels, U.S. Nuclear Regulatory Commission.

-11 NUREG/CR-5704, Effects of LWR Coolant Environment on Fatigue Design Curves of Austenitic Stainless Steel, U.S. Nuclear Regulatory Commission.

-12 10 CFR 50.49, Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants, U. S. Nuclear Regulatory Commission.

-13 Regulatory Guide 1.89, Environmental Qualification of Certain Electrical Equipment Important to Safety for Nuclear Power Plants, U. S. Nuclear Regulatory Commission.

-14 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.

-15 SIR-94-080-A, Relaxation of Reactor Coolant Pump Flywheel Inspection Requirements, Structural Integrity Associates, Inc., September 1997.

15.3-10 Rev. 35

Dome Vertical Horizontal Tendon Dome Tendon Vertical Tendon Horizontal pection Projected Minimum Projected Minimum Projected Minimum Year (kips) Value (kips) (kips) Value (kips) (kips) Value (kips) 40 1453 1343 1521 1339 1467 1325 60 1435 1343 1509 1339 1449 1325 15.3-11 Rev. 35

4.1 ELECTRICAL EQUIPMENT QUALIFICATION 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.

15.4-1 Rev. 35

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.

15.5-1 Rev. 35

le 15.6-1, License Renewal Commitments, provides a listing of the license renewal mitments.

6.1 REFERENCES

FOR SECTION 15.6

-1 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 Number.: 04-398).

-2 Letter from Daniel G. Stoddard to NRC, Dominion Nuclear Connecticut Inc, Millstone Power Station Unit 2, Information in Support of License Renewal Commitment #13, Program Description for Reactor Vessel Internals Inspections, July 31, 2013 (Serial Number 13-398).

-3 Letter from Daniel G. Stoddard to NRC, Dominion Nuclear Connecticut Inc, Millstone Power Station Unit 2, Information in Support of License Renewal Commitment #14, Program Description for Alloy 600 Aging Management Program, July 31, 2013 (Serial Number 13-399).

-4 Letter S/N 05-307, Millstone Power Station Unit 2, Proposed Revision to Technical Specifications (LBDCR 05-MP2-003) Reactor Coolant System Heatup/Cooldown Limits, July 14, 2005.

-5 Letter S/N 06-420, U.S. Nuclear Regulatory Commission, Millstone Power Station, Unit 2, - Issuance of Amendment Re: Revision to Technical Specifications Pertaining to the Reactor Coolant System Heatup and Cooldown Limits (TAC No. MC7593)

May 3, 2006.

15.6-1 Rev. 35

Item Commitment Source Schedule a 1 The existing inspection program will be modified to include those Battery Rack Inspections Complete battery racks 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 Battery Rack Inspections Complete 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.

3 A baseline inspection of the in-scope buried piping located in a damp Buried Pipe Inspection Program Complete 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.

4 The maintenance and work control procedures will be revised to Buried Pipe Inspection Program Complete 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.

5 The Electrical Cables and Connectors Not Subject to 10 CFR 50.49 Electrical Cables and Connectors Complete Environmental Qualification Requirements program will be Not Subject to 10 CFR 50.49 established. Environmental Qualification Requirements 15.6-2 Rev

Item Commitment Source Schedule a 6 Fuse holders meeting the requirements will be evaluated prior to the Electrical Cables and Connectors Complete period of extended operation for possible aging effects requiring Not Subject to 10 CFR 50.49 management. The fuse holder will either be replaced, modified to Environmental Qualification minimize the aging effects, or this program will manage the aging Requirements effects. The program (if needed for fuse holders) will consider the aging stressors for the metallic clips.

7 Procedures will be developed to employ an alternate testing Electrical Cables Not Subject to Prior to Period methodology to confirm the condition of cables and connectors in 10 CFR 50.49 Environmental Extended circuits that have sensitive, low level signals and where the Qualification Requirements Used in Operation.

instrumentation is not calibrated in situ. Instrumentation Circuits Complete.

Subsequent test will not exceed 10 year frequen 8 A baseline visual inspection will be performed on a representative Fire Protection Program Complete sample of 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 Fire Protection Program Prior to The replacing those that have been in service for 50 years has been Sprinkler Heads included in Fire Protection Program. Achieving 50 Years Of Servic Life Not to Exceed a Year Interval Thereafter 10 The procedures and training for personnel performing General General Condition Monitoring Complete Condition Monitoring inspections and walkdowns will be enhanced to provide expectations that identify the requirements for the inspection of aging effects.

15.6-3 Rev

Item Commitment Source Schedule a 11 In-scope cable found to be submerged will be subject to an Inaccessible Medium Voltage Prior to period engineering evaluation and corrective action. The evaluation of cables Cables Not Subject to 10 CFR 50.49 extended having significant voltage found to be submerged in standing water Environmental Qualification operation.

for an extended period of time will be based on appropriate testing Requirements Complete.

(using available technology consistent with NRC positions) of cables Subsequent test that are determined to be wetted for a significant period of time. The will not exceed Engineering evaluation will also address the appropriate testing 10 year frequen 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.

12 The Infrequently Accessed Areas Inspection Program will be Infrequently Accessed Areas Complete established. Inspection Program 13 Millstone will follow the industry efforts on reactor vessel internals ISI Program: Reactor Vessel Complete regarding such issues as thermal or neutron irradiation embrittlement Internals (Reference 15.6-2)

(loss of fracture toughness), void swelling (change in dimensions),

and stress corrosion cracking (PWSCC and IASCC) 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.

15.6-4 Rev

Item Commitment Source Schedule a 14 Millstone will follow the industry efforts investigating the aging Alloy 600 Management Program At Least Two effects applicable to nickel-based alloys (i.e., PWSCC in Alloy 600 Years Prior to th base metal and Alloy 82/182 weld metals) and identifying the Period of Exten appropriate aging management activities and will implement the Operation appropriate recommendations resulting from this guidance. The revised program description will be submitted at least two years prior Complete 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 54.21(a)(3).

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

16 Implementing procedures and documentation will be modified to Inspection Activities: Load Complete include visual inspections for the loss of material on the crane and Handling Cranes and Devices trolley structural components and the rails in the scope of license renewal in Commitment 15.

17 The implementing procedures will be modified to include ACI Structures Monitoring Program Complete 349.3R-96 and ANSI/ASCE 11-90 as references and as input documents for the inspection program.

18 The Structures Monitoring Program and implementing procedures Structures Monitoring Program Complete will be modified to include all in-scope structures.

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

20 The Structures Monitoring Program and implementing procedures Structures Monitoring Program Complete 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.

15.6-5 Rev

Item Commitment Source Schedule a 21 The maintenance and work control procedures will be revised to Structures Monitoring Program Complete 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.

22 Appropriate inspections of sealants and caulkings used for moisture Tank Inspection Program Complete intrusion prevention in and around aboveground tanks will be performed.

23 Non-destructive volumetric examination of the in-scope inaccessible Tank Inspection Program Applicable locations, such as the external surfaces of tank bottoms, will be inspections hav performed prior to the period of extended operation. Subsequent been performed inspections will be performed on a frequency consistent with Subsequent scheduled tank internals inspection activities. inspections will performed on a frequency consistent with scheduled tank internal inspecti activities.

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

25 Changes will be made to maintenance and work control procedures to Work Control Process Complete ensure that inspections of plant components and plant commodities will be appropriately and consistently performed and documented for aging effects during maintenance activities.

15.6-6 Rev

Item Commitment Source Schedule a 26 Dominion actively participates in a comprehensive industry initiative, Environmental Report - SAMA Complete in response to NRC Generic Issue 23 (GI-23), Reactor Coolant Analysis Pump Seal Failure. Dominion is following the industry efforts on this issue and will implement the appropriate recommendations resulting from this guidance prior to the period of extended operation.

27 For potentially susceptible CASS materials, either enhanced Inservice Inspection Program: Complete volumetric examinations or a unit or component specific flaw Systems, Components and Supports 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.

28 Millstone will follow industry efforts that will provide specific Environmentally Assisted Fatigue Complete 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.

29 A baseline visual inspection will be performed of the accessible areas Closed-Cycle Cooling Water Complete of the shell side (including accessible portions of the exterior side of System the tubes) of one:

  • Millstone Unit 2 Reactor Building Closed Cooling Water heat exchanger,

15.6-7 Rev

Item Commitment Source Schedule a 30 Using the Work Control Process, a baseline inspection for the loss of Work Control Process Complete 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.

31 A review of the Work Control Process inspection opportunities for Work Control Process Complete each 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.

32 Calibration results for cables tested in situ will be reviewed to detect Electrical Cables Not Subject to Prior to period o severe aging degradation of the cable insulation. The initial review 10 CFR 50.49 Environmental extended operat will be completed prior to entering the period of extended operation Qualification Requirements Used in complete.

and will include at least 5 years of surveillance test data for each cable Instrumentation Circuits Subsequent reviewed. Subsequent reviews will be performed on a period not to reviews will not exceed 10 years. exceed a 10 yea frequency.

33 The in scope cables in Unit 3 duct lines number 929 (SBO Diesel to Inaccessible Medium Voltage Prior to period Unit 3 4.16kV Normal Switchgear) and number 973 (RSST 3RTX- Cables Not Subject to 10 CFR 50.49 extended XSR-B to 6.9kV Normal Switchgear Bus 35A, 35B, 35C and 35D) Environmental Qualification operation.

will be tested to demonstrate that water treeing will not prevent the Requirements Complete.

cables from performing their intended function. Subsequent test will not exceed 10 year frequen 15.6-8 Rev

Item Commitment Source Schedule a 34 In addition to the testing specified in Commitment 33, a Inaccessible Medium Voltage Prior to Period representative sample of in-scope medium voltage cables will be Cables Not Subject to 10 CFR 50.49 Extended tested to demonstrate that water treeing will not prevent the cables Environmental Qualification Operation from performing their intended function. Requirements Not to Exceed a Year Frequency Thereafter 35 Millstone Unit 2 will complete the SAMA evaluation of the capability Severe Accident Mitigation Complete to flash the Diesel Generator field in the event of extended loss of DC Alternatives (SAMA) Analysis power with a loss of offsite power. If this SAMA is cost beneficial (Reference 15.6-1)

(i.e., can be accomplished without a hardware modification), a Severe Accident Management Guideline (SAMG) addressing this mitigation strategy will be developed.

36 Dominion will replace the Millstone Unit 2 pressurizer using Inservice Inspection Program: Complete materials that are resistant to PWSCC. Systems, Components and Supports 37 Updated USE, RTPTS, and P-T limits based on fluence values Reactor Vessel Neutron Prior to period developed in accordance with Regulatory Guide 1.190 requirements, Embrittlement extended as amended or superseded by future regulatory guidance changes, will operation.

be submitted to the NRC for review at least two years prior to the Complete.

period of extended operation. Subsequent test will not exceed 10 year frequen

a. The Period of Extended Operation is the period of 20 years beyond the expiration date of the units previous 40 year operating lice 15.6-9 Rev
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