SBK-L-17133, Clarification to Supplement 55 - Response to Request for Additional Information for the Review of the Seabrook Station License Renewal Application - Bolting Integrity Program

Clarification to Supplement 55 - Response to Request for Additional Information for the Review of the Seabrook Station License Renewal Application - Bolting Integrity Program

ML17223A022
Person / Time
Site:	Seabrook
Issue date:	08/11/2017
From:	Mccartney E NextEra Energy Seabrook
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SBK-L-17133
Download: ML17223A022 (30)

Text

August 11, 2017 10CFR54 Docket No. 50-443 SBK-L-17133 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Seabrook Station NEX era ~

EN ERGY ~

SEABROOK Clarification to Supplement 55 - Response to Request for Additional Information for the Review of the Seabrook Station License Renewal Application - Bolting Integrity Program

References:

1. NextEra Energy Seabrook LLC, letter SBK-L-10077, "Seabrook Station Application for Renewed Operating License," May 25, 2010 (Accession Number ML101590099).

2. NextEra Energy Seabrook LLC, letter SBK-L-12258, "LRA Commitment Update and RAI B.1.4-4 Clarification," December 10, 2012 (Accession Number ML12349A214).

3. NRC, "Final Request for Additional Information Regarding Seabrook License Renewal Application - Set 26 (CAC NO. ME4028), RAI B.2.1.9 Bolting Integrity,"

May 24, 2017 (Accession Number ML17139B835).

4. NextEra Energy Seabrook LLC, Letter SBK-L-17103, "Supplement 55 - Response to Request for Additional Information for the Review of Seabrook Station License Renewal Application - Bolting Integrity Program," June 20, 2017 (Accession Number ML17173A065).

In Reference 1, NextEra Energy Seabrook, LLC (NextEra Energy Seabrook) submitted an application for a renewed facility operating license for Seabrook Station Unit 1 in NextEra Energy Seabrook, LLC P.O. Box 300, Lafayette Road, Seabrook, NH 03874

U.S. Nuclear Regulatory Commission SBK-L-17133 I Page 2 accordance with the Code of Federal Regulations, Title 10, Parts 50, 51, and 54.

In Reference 2, NextEra Energy Seabrook submitted letter SBK-L-12258, which revised License Renewal Application Appendix A, A.2.1.9, and Appendix B, B.2.1.9.

In Reference 3, the NRC requested additional information (RAI) related to NextEra Energy Seabrook's Bolting Integrity Program and how the aging effects for closure bolting associated with air filled, gas filled, and systems at atmospheric pressure will be managed.

In Reference 4, NextEra Energy Seabrook submitted letter SBK-L-17103, responding to the Request for Additional Information in Reference 3.

On July 11, 2017, the NRC Staff and NextEra Energy Seabrook had a clarification call discussing the response (SBK-L-17103) to the Bolting Integrity RAI set in Reference 3. of this letter clarifies the RAI response provided within SBK-L-17103. provides the revised License Renewal Application (LRA), Appendix B -

Aging Management Programs, B.2.1.9 - Bolting Integrity.

To facilitate understanding, the changes are explained, and where appropriate, portions of the LRA are repeated with the change highlighted by strikethroughs for deleted text and balded italics for inserted text.

This letter contains one new Commitment, #51. Enclosure 3 provides the revised LRA Appendix A - Updated Final Safety Analysis Report Supplement Table A.3, License Renewal Commitment List.

If there are any questions or additional information is needed, please contact Mr.

Edward J. Carley, Engineering Supervisor - License Renewal, at (603) 773-7957.

If you have any questions regarding this correspondence, please contact Mr. Kenneth Browne, Licensing Manager, at (603) 773-7932.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on August IQ

, 2017.

Sincerely, NextEra Energy Seabrook, LLC

¥:vz~~

U.S. Nuclear Regulatory Commission SBK-L-17133 I Page 3 : Clarification to Supplement 55 - Response to Request for Additional Information for the Review of the Seabrook Station License Renewal Application - Bolting Integrity Program : Revised License Renewal Application, Appendix B - Aging Management Programs, B.2.1.9 - Bolting Integrity : Revised LRA Appendix A - Updated Final Safety Analysis Report Supplement Table A.3, License Renewal Commitment List cc:

D. H. Dorman J.C. Poole P. C. Cataldo L. M. James Mr. Perry Plummer NRC Region I Administrator NRC Project Manager NRC Senior Resident Inspector NRC Project Manager, License Renewal Director Homeland Security and Emergency Management New Hampshire Department of Safety Division of Homeland Security and Emergency Management Bureau of Emergency Management 33 Hazen Drive Concord, NH 03305 perry.plummer@dos.nh.gov Mr. John Giarrusso, Jr., Nuclear Preparedness Manager The Commonwealth of Massachusetts Emergency Management Agency 400 Worcester Road Framingham, MA 01702-5399 John. Giarrusso@massmail.state. ma. us to SBK-L-17133 Clarification to Supplement 55 - Response to Request for Additional Information for the Review of the Seabrook Station License Renewal Application - Bolting Integrity Program

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure 1IPage2 On June 20, 2017, NextEra Energy Seabrook submitted letter SBK-L-17103 to supplement the License Renewal Application. The letter responded to a Request for Additional Information (RAI) on the station's License Renewal Bolting Integrity Program and how the aging effects for closure bolting associated with air filled, gas filled, and systems at atmospheric pressure will be managed. On July 11, 2017, the Staff and NextEra Energy Seabrook had a clarification call discussing the response to the RAI set. The response provided herein supersedes the previous response provided in SBK-L-17103.

RAI # B.2.1.9-3

Background

Section 54.21(a)(3) of 10 CFR requires the applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. As described in SRP-LR, an applicant may demonstrate compliance with 10 CFR 54.21(a)(3) by referencing the GALL Report and when evaluation of the matter in the GALL Report applies to the plant.

GALL Report AMP Xl.M18 states:

Degradation of pressure boundary closure bolting due to crack initiation, loss of preload, or loss of material may result in leakage from the mating surfaces or joint connections of pressure boundary components. Periodic inspection of pressure boundary components for signs of leakage ensures that age-related degradation of closure bolting is detected and corrected before component leakage becomes excessive. Accordingly, pressure retaining bolted connections should be inspected at least once per refueling cycle.

LRA Section B.2.1.9 states that the Bolting Integrity program "includes periodic inspection of closure bolting assemblies to detect signs of leakage that may be indicative of loss of preload...

Periodic inspection of bolted closures in conjunction with the Seabrook Station In Service Inspection Program and Seabrook Station External Surfaces Monitoring Program will detect the aging effects and joint leakage." The LRA states that the External Surfaces Monitoring and In Service Inspection Programs use visual inspections to detect leakage which could indicate a loss of preload for closure bolting.

Issue While the staff notes that visual inspections will be capable of detecting leakage of liquid filled systems, a visual inspection may not be sufficient to detect leakage for air-or gas filled systems.

LRA Table 3.3.2-8, "Containment Online Purge System," and 3.3.2-20, "Instrument Air System,"

are examples of systems that cite the Bolting Integrity Program to manage aging effects associated with closure bolting that could be or are air-filled or gas-filled. Based on a review of LRA Section B.2.1.9 and SER with Open Items Section 3.0.3.1.7, "Bolting Integrity," it is not clear how aging effects for closure bolting installed in air-filled or gas-filled systems will be effectively managed because in these systems there is typically limited or no visual signs of leakage.

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure I I Page 3 NRC Request #1 For air-filled systems, gas-filled systems, and systems at internal atmosphere pressure for which aging effects for closure bolting will be managed by the Bolting Integrity Program, state how the aging effects will be managed.

NextEra Energy Seabrook's Response to RAI B.2.1.9-3, Request #1 Seabrook will manage aging effects for the Bolting Integrity Program closure bolting in air and gas filled systems by using an applicable inspection technique that ensures the integrity of bolted joints will be demonstrated. These inspection techniques will be one of the following:

Visual inspection for discoloration when the inside environment would discolor the external surface of the pipe if leakage would occur.

Monitoring and trending of pressure decay when the bolted connection is within an isolable boundary.

Soap like solution to promote surface bubbling.

Thermography when the piping's internal fluid temperature is higher than the ambient environment.

Operations personnel will identify and monitor leaks as part of their routine walk downs. Plant Engineering also perform system walkdowns which also provides an opportunity to discover bolted joint leakage.

For Bolting Integrity Program closure bolting associated with systems at internal atmospheric pressure, Seabrook will perform a tightness check on 20 percent of in scope bolts with a maximum of 25 bolts per population. Populations will be of the same material and environment combination. Inspections will occur before the period of extended operation (PEO}, and then every 10 years after the initial inspection date. By checking the tightness of these closure bolting samples, the inspections verify that potential loss of material, loss of preload, or cracking is not occurring.

The revised License Renewal Application, Appendix B -Aging Management Programs, B.2.1.9

- Bolting Integrity, Program Description and Enhancements sections, are provided within of this letter.

NRC Request #2 State any enhancements to plant-specific procedures and/or to the Bolting Integrity Program necessary to incorporate how these aging effects will be managed.

NextEra Energy's Seabrook Response to RAI B.2.1.9-3, Request #2 The revised License Renewal Application, Appendix B - Aging Management Programs, B.2.1.9

- Bolting Integrity, Program Description and Enhancements sections, are provided within of this letter.

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure 1IPage4 Commitment #51 is used to track the enhancements provided in Enclosure 2, and Enclosure 3 provides the revised LRA Appendix A - Updated Final Safety Analysis Report Supplement Table A.3, License Renewal Commitment List.

E1th{[1tce the program to lll{[lt([ge the

{[gi11g effects for closure bolti11g wit/tin (fir mu/ g{[s filled systems by using (Ill

{[pp/iC(lb/e i11spectio11 teclt11ique tlwt e11sures the integrity of bolted joints will be demonstrated.

For closure bolting wit/tin systems (If Prior to the period 51 Bolting Integrity

{[fmosplteric pressure, tigltt11ess checks A.2.1.9 of extended will be pe1for111ed 011 20 percent of bolts operation.

with {[ 11wximu111 of 25 bolts per pop11/{[fio11. Popu[{[tions will be of the smne 111([/erfol m1d e11viro11me11t combilwtio11. J11spectio11s will occur before the period of extended operation,

{[lid tit en every I 0 ye{[rs <ifter the i11itfol inspection <fote.

to SBK-L-17133 Revised License Renewal Application, Appendix B - Aging Management Programs, B.2.1.9 - Bolting Integrity

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure 2 I Page 2 Enclosure bolting of air filled, gas filled, and systems at atmospheric pressure within the Bolting Integrity program will have the appropriate aging management. As transmitted previously in SBK-L-12258 (Reference 2), and in SBK-L-17103 (Reference 4), the second paragraph of Seabrook Station's License Renewal Application, Appendix B - Aging Management Programs, B.2.1.9 - Bolting Integrity, Program Description, is revised below.

The program includes periodic inspection of closure bolting assemblies to detect signs of leakage that may be indicative of loss of preload, loss of material, or crack initiation. Leak detection inspections are performed on in scope closure bolting in air filled and gas filled systems using a method that detects leakage. Prior to the period of extended operation, readily accessible in scope closure bolting within systems at internal atmospheric pressure will have a tightness check performed. Tightness checks will be perlormed on 20 percent of bolts with a maximum of 25 bolts per population. Populations will be of the same material and environment combination. Inspections will occur before the period of extended operation (PEO), and then every 10 years after the initial inspection date.

Periodic inspections of bolted closures in conjunction with the Seabrook Station lnservice Inspection Program and Seabrook Station External Surfaces Monitoring Program, will detect the aging effects and joint leakage. Operator rounds and system walkdowns will also identify joint leakage. Bolted connections in buried, underground, and inaccessible submerged piping will be inspected for indication of leakage caused by loss of preload when the associated piping is inspected by the Buried Piping and Tanks Inspection program.

In instances where hydrostatic testing, flow testing, or fire protection jockey pump monitoring are used in lieu of visual inspections, these methods will also be credited to identify leakage caused by loss of preload at bolted connections. Additionally, the service water pump column bolted connections are submerged in raw water and cannot be inspected for leakage due to loss of preload. Service water pump column bolting will be inspected for loss of material and loss of preload when the service water pumps are removed for maintenance.

The Seabrook Station's License Renewal Application, Appendix B - Aging Management Programs, B.2.1.9 - Bolting Integrity, Enhancements, is revised below. Commitment #51 is used to track the enhancements.

Enhancements The following enhancements will be made prior to entering the period of extended operation.

Seabrook will manage aging effects for the Bolting Integrity Program closure bolting in air and gas filled systems by using an applicable inspection technique that ensures the integrity of bolted joints will be demonstrated. These inspection techniques will be one of the following:

a.

Visual inspection for discoloration when the inside environment would discolor the external surlace of the pipe if leakage would occur.

b.

Monitoring and trending of pressure decay when the bolted connection is within an isolable boundary.

c.

Soap like solution to promote surlace bubbling.

d.

Thermography when the piping's internal fluid is higher than the ambient environment.

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure 2 I Page 3 Closure bolting associated with systems at internal atmospheric pressure will have a tightness check performed. Tightness checks will be performed on 20 percent of bolts with a maximum of 25 bolts, per population. Populations will be of the same material and environment combination. Inspections will occur before the period of extended operation (PEO), and then every 10 years after the initial inspection date.

Program Elements Affected: Element 1 (Scope of Program), Element 4 (Detection of Aging Effects).

to SBK-L-17133 Revised LRA Appendix A - Updated Final Safety Analysis Report Supplement Table A.3, License Renewal Commitment List

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure 3 I Page 2 A.3 LICENSE RENEW AL COMMITMENT LIST No.

PROGRAM or TOPIC COMMITMENT Provide confirmation and acceptability of the I.

PWR Vessel Internals implementation ofMRP-227-A by addressing the plant-specific Applicant/Licensee Action Items outlined in section 4.2 of the NRC SER.

Enhance the program to include visual inspection for

2.

Closed-Cycle Cooling Water cracking, loss of material and fouling when the in-scope systems are opened for maintenance.

Inspection of Overhead Heavy Enhance the program to monitor general corrosion on the

3.

Load and Light Load (Related to crane and trolley structural components and the effects of Refueling) Handling Systems wear on the rails in the rail system.

Inspection of Overhead Heavy Enhance the program to list additional cranes for

4.

Load and Light Load (Related to Refueling) Handling Systems monitoring.

Enhance the program to include an annual air quality test

5.

Compressed Air Monitoring requirement for the Diesel Generator compressed air sub system.

6.

Fire Protection Enhance the program to perform visual inspection of penetration seals by a fire protection qualified inspector.

Enhance the program to add inspection requirements such

7.

Fire Protection as spalling, and loss of material caused by freeze-thaw, chemical attack, and reaction with aggregates by qualified inspector.

UFSAR SCHEDULE LOCATION A.2.1.7 Complete Prior to the period of A.2.1.12 extended operation.

Prior to the period of A.2.1.13 extended operation.

Prior to the period of A.2.1.13 extended operation.

Prior to the period of A.2.1.14 extended operation.

A.2.1.15 Prior to the period of extended operation.

A.2.1.15 Prior to the period of extended operation.

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure 3 I Page 3

8.

Fire Protection

9.

Fire Water System

10.

Fire Water System

11.

Fire Water System Enhance the program to include the performance of visual inspection of fire-rated doors by a fire protection qualified inspector.

Enhance the program to include NFPA 25 (2011 Edition) guidance for "where sprinklers have been in place for 50 years, they shall be replaced or representative samples from one or more sample areas shall be submitted to a recognized testing laboratory for field service testing".

Enhance the program to include the performance of periodic flow testing of the fire water system in accordance with the guidance ofNFP A 25 (2011 Edition).

Enhance the program to include the performance of periodic visual or volumetric inspection of the internal surface of the fire protection system upon each entry to the system for routine or corrective maintenance to evaluate wall thickness and inner diameter of the fire protection piping ensuring that corrosion product buildup will not result in flow blockage due to fouling. Where surface irregularities are detected, follow-up volumetric examinations are performed. These inspections will be documented and trended to determine if a representative number of inspections have been performed prior to the period of extended operation. If a representative number of inspections have not been performed prior to the period of extended operation, focused inspections will be conducted. These inspections will commence during the ten year period prior to the period of extended operation and continue through the period of extended operation.

Prior to the period of A.2.1.15 extended operation.

Prior to the period of A.2.1.16 extended operation.

Prior to the period of A.2.1.16 extended operation.

Within ten years prior to the A.2.1.16 period of extended operation.

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure 3 I Page 4

12.

Aboveground Steel Tanks

13.

Fire Water System

14.

Fuel Oil Chemistry

15.

Fuel Oil Chemistry Enhance the program to include 1) In-scope outdoor tanks, except fire water storage tanks, constructed on soil or concrete, 2) Indoor large volume storage tanks (greater than 100,000 gallons) designed to near-atmospheric internal pressures, sit on concrete or soil, and exposed internally to water, 3) Visual, surface, and volumetric examinations of the outside and inside surfaces for managing the aging effects of loss of material and cracking,

4) External visual examinations to monitor degradation of the protective paint or coating, and 5) Inspection of sealant and caulking for degradation by performing visual and tactile examination (manual manipulation) consisting of pressing on the sealant or caulking to detect a reduction in the resiliency and pliability.

Enhance the program to perform exterior inspection of the fire water storage tanks annually for signs of degradation and include an ultrasonic inspection and evaluation of the internal bottom surface of the two Fire Protection Water Storage Tanks per the guidance provided in NFP A 25 (2011 Edition).

Enhance program to add requirements to 1) sample and analyze new fuel deliveries for biodiesel prior to offloading to the Auxiliary Boiler fuel oil storage tank and

2) periodically sample stored fuel in the Auxiliary Boiler fuel oil storage tank.

Enhance the program to add requirements to check for the presence of water in the Auxiliary Boiler fuel oil storage tank at least once per quarter and to remove water as necessary.

A.2.1.17 Within 10 years prior to the period of extended operation.

A.2.1.16 Within ten years prior to the period of extended operation.

Prior to the period of A.2.1.18 extended operation.

A.2.1.18 Prior to the period of extended operation.

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure 3 I Page 5

16.

Fuel Oil Chemistry

17.

Fuel Oil Chemistry

18.

Reactor Vessel Surveillance

19.

Reactor Vessel Surveillance

20.

Reactor Vessel Surveillance

21.

Reactor Vessel Surveillance Enhance the program to require draining, cleaning and inspection of the diesel fire pump fuel oil day tanks on a frequency of at least once every ten years.

Enhance the program to require ultrasonic thickness measurement of the tank bottom during the IO-year draining, cleaning and inspection of the Diesel Generator fuel oil storage tanks, Diesel Generator fuel oil day tanks, diesel fire pump fuel oil day tanks and auxiliary boiler fuel oil storage tank.

Enhance the program to specify that all pulled and tested capsules, unless discarded before August 31, 2000, are placed in storage.

Enhance the program to specify that if plant operations exceed the limitations or bounds defined by the Reactor Vessel Surveillance Program, such as operating at a lower cold leg temperature or higher fluence, the impact of plant operation changes on the extent of Reactor Vessel embrittlement will be evaluated and the NRC will be notified.

Enhance the program as necessary to ensure the appropriate withdrawal schedule for capsules remaining in the vessel such that one capsule will be withdrawn at an outage in which the capsule receives a neutron fluence that meets the schedule requirements of 10 CFR 50 Appendix H and ASTM E 185-82 and that bounds the 60-year fluence, and the remaining capsule(s) will be removed from the vessel unless determined to provide meaningful metallurgical data.

Enhance the program to ensure that any capsule removed, without the intent to test it, is stored in a manner which maintains it in a condition which would permit its future use, including during the period of extended operation.

Prior to the period of A.2.1.18 extended operation.

A.2.1.18 Prior to the period of extended operation.

Prior to the period of A.2.1.19 extended operation.

Prior to the period of A.2.1.19 extended operation.

Prior to the period of A.2.1.19 extended operation.

A.2.1.19 Prior to the period of extended operation.

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure 3 I Page 6

22.

One-Time Inspection 1~

_.)_

Selective Leaching of Materials Buried Piping And Tanks

24.

Inspection

25.

One-Time Inspection of ASME Code Class 1 Small Bore-Piping

26.

External Surfaces Monitoring Inspection of Internal Surfaces in

27.

Miscellaneous Piping and Ducting Components

28.

Lubricating Oil Analysis

29.

Lubricating Oil Analysis

30.

Lubricating Oil Analysis Implement the One Time Inspection Program.

A.2.1.20 Within ten years prior to the period of extended operation.

Implement the Selective Leaching of Materials Program.

The program will include a one-time inspection of Within five years prior to the selected components where selective leaching has not A.2.1.21 period of extended operation.

been identified and periodic inspections of selected components where selective leaching has been identified.

Implement the Buried Piping And Tanks Inspection Within ten years prior to the Program.

A.2.1.22 period of extended operation.

Implement the One-Time Inspection of ASME Code Class A.2.1.23 Within ten years prior to the I Small Bore-Piping Program.

period of extended operation.

Enhance the program to specifically address the scope of the program, relevant degradation mechanisms and effects Prior to the period of of interest, the refueling outage inspection frequency, the A.2.1.24 training requirements for inspectors and the required extended operation.

periodic reviews to determine program effectiveness.

Implement the Inspection oflnternal Surfaces in Prior to the period of Miscellaneous Piping and Ducting Components Program.

A.2.1.25 extended operation.

Enhance the program to add required equipment, lube oil Prior to the period of analysis required, sampling frequency, and periodic oil A.2.1.26 changes.

extended operation.

Enhance the program to sample the oil for the Reactor A.2.1.26 Prior to the period of Coolant pump oil collection tanks.

extended operation.

Enhance the program to require the performance of a one-time ultrasonic thickness measurement of the lower A.2.1.26 Prior to the period of portion of the Reactor Coolant pump oil collection tanks extended operation.

prior to the period of extended operation.

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure 3 I Page 7 ASME Section XI, Subsection

31.

IWL

32.

Structures Monitoring Program

.).).

Structures Monitoring Program Electrical Cables and

34.

Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Electrical Cables and Connections Not Subject to 10

35.

CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits Inaccessible Power Cables Not Subject to IO CFR 50.49

36.

Environmental Qualification Requirements

37.

Metal Enclosed Bus

38.

Fuse Holders Electrical Cable Connections Not Subject to IO CFR 50.49

39.

Environmental Qualification Requirements Enhance procedure to include the definition of A.2.1.28 Prior to the period of "Responsible Engineer".

extended operation.

Enhance procedure to add the aging effects, additional Prior to the period of locations, inspection frequency and ultrasonic test A.2.1.31 extended operation.

requirements.

Enhance procedure to include inspection of opportunity Prior to the period of when planning excavation work that would expose A.2.1.31 extended operation.

inaccessible concrete.

Implement the Electrical Cables and Connections Not Prior to the period of Subject to 10 CFR 50.49 Environmental Qualification A.2.1.32 extended operation.

Requirements program.

Implement the Electrical Cables and Connections Not Prior to the period of Subject to 10 CFR 50.49 Environmental Qualification A.2.1.33 extended operation.

Requirements Used in Instrumentation Circuits program.

Implement the Inaccessible Power Cables Not Subject to Prior to the period of IO CFR 50.49 Environmental Qualification Requirements A.2.1.34 extended operation.

program.

Implement the Metal Enclosed Bus program.

A.2.1.35 Prior to the period of extended operation.

Implement the Fuse Holders program.

A.2.1.36 Prior to the period of extended operation.

Implement the Electrical Cable Connections Not Subject Prior to the period of to 10 CFR 50.49 Environmental Qualification A.2.1.37 extended operation.

Requirements program.

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure 3 I Page 8

40.

345 KV SF6 Bus Metal Fatigue of Reactor Coolant

41.

Pressure Boundary Metal Fatigue of Reactor Coolant

42.

Pressure Boundary Pressure -Temperature Limits,

43.

including Low Temperature Overpressure Protection Limits Implement the 345 KV SF6 Bus program.

Enhance the program to include additional transients beyond those defined in the Technical Specifications and UFSAR.

Enhance the program to implement a software program, to count transients to monitor cumulative usage on selected components.

Seabrook Station will submit updates to the P-T curves and LTOP limits to the NRC at the appropriate time to comply with 10 CFR 50 Appendix G.

A.2.2.1 Prior to the period of extended operation.

Prior to the period of A.2.3.1 extended operation.

Prior to the period of A.2.3.1 extended operation.

The updated analyses will be submitted at the appropriate A.2.4.1.4 time to comply with I 0 CFR 50 Appendix G, Fracture Toughness Requirements.

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure 3 I Page 9

44.

Environmentally-Assisted Fatigue Analyses (TLAA)

NextEra Seabrook will perform a review of design basis ASME Class 1 component fatigue evaluations to determine whether the NUREG/CR-6260-based components that have been evaluated for the effects of the reactor coolant environment on fatigue usage are the limiting components for the Seabrook plant configuration. If more limiting components are identified, the most limiting component will be evaluated for the effects of the reactor coolant environment on fatigue usage. If the limiting location identified consists of nickel alloy, the environmentally-assisted fatigue calculation for nickel alloy will be performed using the rules ofNUREG/CR-6909.

(1) Consistent with the Metal Fatigue of Reactor Coolant Pressure Boundary Program Seabrook Station will update the fatigue usage calculations using refined fatigue analyses, if necessary, to determine acceptable CUFs (i.e., less than 1.0) when accounting for the effects of the reactor water environment. This includes applying the appropriate Fen factors A.2.4.2.3 At least two years prior to the to valid CUFs determined from an existing fatigue analysis valid period of extended operation.

for the period of extended operation or from an analysis using an NRC-approved version of the ASME code or NRC-approved alternative (e.g., NRC-approved code case).

(2) If acceptable CUFs cannot be demonstrated for all the selected locations, then additional plant-specific locations will be evaluated. For the additional plant-specific locations, ifCUF, including environmental effects is greater than 1.0, then Corrective Actions will be initiated, in accordance with the Metal Fatigue of Reactor Coolant Pressure Boundary Program, B.2.3.1. Corrective Actions will include inspection, repair, or replacement of the affected locations before exceeding a CUF of 1.0 or the effects of fatigue will be managed by an inspection program that has been reviewed and approved by the NRC (e.g.,

periodic non-destructive examination of the affected locations at inspection intervals to be determined by a method accepted by theNRC).

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure 3 I Page 10

45.

Alkali-Silica Reaction (ASR)

Monitoring Program Protective Coating Monitoring

46.

and Maintenance Protective Coating Monitoring

47.

and Maintenance

48.

Protective Coating Monitoring and Maintenance

49.

Protective Coating Monitoring and Maintenance ASME Section XI, Subsection

50.

IWE NextEra will obtain additional cores in the vicinity of three extensometers and perform modulus testing. Using these test results, NextEra will determine the change in through-thickness expansion since installation of the extensometers and compare it to change determined from extensometer readings. Consistency between these results will provide additional corroboration of the methodology in MPR-4153.

Enhance the program by designating and qualifying an Inspector Coordinator and an Inspection Results Evaluator.

Enhance the program by including, "Instruments and Equipment needed for inspection may include, but not be limited to, flashlight, spotlights, marker pen, mirror, measuring tape, magnifier, binoculars, camera with or without wide angle lens, and self sealing polyethylene sample bags."

Enhance the program to include a review of the previous two monitoring reports.

Enhance the program to require that the inspection report is to be evaluated by the responsible evaluation personnel, who is to prepare a summary of findings and recommendations for future surveillance or repair.

Perform UT of the accessible areas of the containment liner plate in the vicinity of the moisture barrier for loss of material. Perform opportunistic UT of inaccessible areas.

A.2.1.31.A At least 2 years prior to the period of extended operation.

Prior to the period of A.2.1.38 extended operation.

A.2.1.38 Prior to the period of extended operation.

A.2.1.38 Prior to the period of extended operation.

A.2.1.38 Prior to the period of extended operation.

Baseline inspections were completed during ORI 6.

A.2.1.27 Repeat containment liner UT thickness examinations at intervals of no more than five (5) refueling outages.

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure 3 I Page 11

51.

Bolting Integrity ASME Section XI, Subsection

52.

IWL

53.

Reactor Head Closure Studs Enhance the program to manage the aging effects for closure bolting within air and gas filled systems by using an applicable inspection technique that ensures the integrity of bolted joints will be demonstrated.

For closure bolting within systems at atmospheric Prior to the period of pressure, tightness checks will be peiformed 011 20 A.2.1.9 extended operation.

percent of bolts with a maximum of 25 bolts per population. Populations will be of the same material and environment combination. Inspections will occur before the period of extended operation, and then every 10 years after the initial inspection date.

Implement measures to maintain the exterior surface of the Containment Structure, from elevation -30 feet to +20 A.2.1.28 Complete feet, in a dewatered state.

Replace the spare reactor head closure stud(s)

Prior to the period of manufactured from the bar that has a yield strength> 150 A.2.1.3 ksi with ones that do not exceed 150 ksi.

extended operation.

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure 3 I Page 12

54.

Steam Generator Tube Integrity

55.

Number Not Used Closed-Cycle Cooling Water

56.

System Closed-Cycle Cooling Water

57.

System NextEra will address the potential for cracking of the primary to secondary pressure boundary due to PWSCC oftube-to-tubesheet welds using one of the following two options:

I) Perform a one-time inspection of a representative sample oftube-to-tubesheet welds in all steam generators to determine if PWSCC cracking is present and, if cracking is identified, resolve the condition through engineering evaluation justifying continued operation or repair the condition, as appropriate, and establish an ongoing monitoring program to perform routine tube-to-tubesheet weld inspections for the remaining life of the A.2.1.10 Complete steam generators, or

2) Perform an analytical evaluation showing that the structural integrity of the steam generator tube-to-tubesheet interface is adequately maintaining the pressure boundary in the presence oftube-to-tubesheet weld cracking, or redefining the pressure boundary in which the tube-to-tubesheet weld is no longer included and, therefore, is not required for reactor coolant pressure boundary function. The redefinition of the reactor coolant pressure boundary must be approved by the NRC as part of a license amendment request.

Revise the station program documents to reflect the EPRI Prior to the period of Guideline operating ranges and Action Level values for A.2.1.12 hydrazine and sulfates.

extended operation.

Revise the station program documents to reflect the EPRI Prior to the period of Guideline operating ranges and Action Level values for A.2.1.12 Diesel Generator Cooling Water Jacket pH.

extended operation.

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure 3 I Page 13

58.

Fuel Oil Chemistry Nickel Alloy Nozzles and

59.

Penetrations Buried Piping and Tanks

60.

Inspection Compressed Air Monitoring

61.

Program

62.

Water Chemistry

63.

Flow Induced Erosion

64.

Buried Piping and Tanks Inspection Update Technical Requirement Program 5.1, (Diesel Fuel Oil Testing Program) ASTM standards to ASTM D2709-A.2.1.18 Prior to the period of 96 and ASTM D4057-95 required by the GALL XI.M30 extended operation.

Rev 1.

The Nickel Alloy Aging Nozzles and Penetrations Prior to the period of program will implement applicable Bulletins, Generic A.2.2.3 extended operation.

Letters, and staff accepted industry guidelines.

Implement the design change replacing the buried Prior to the period of Auxiliary Boiler supply piping with a pipe-within-pipe A.2.1.22 configuration with leak detection capability.

extended operation.

Replace the flexible hoses associated with the Diesel Within ten years prior to the Generator air compressors on a frequency of every 10 A.2.1.14 period of extended operation.

years.

Enhance the program to include a statement that sampling Prior to the period of frequencies are increased when chemistry action levels are A.2.1.2 exceeded.

extended operation.

Ensure that the quarterly CVCS Charging Pump testing is continued during the PEO. Additionally, add a precaution to the test procedure to state that an increase in the eves Prior to the period of Charging Pump mini flow above the acceptance criteria A.2.1.2 extended operation.

may be indicative of erosion of the mini flow orifice as described in LER 50-275/94-023.

Soil analysis shall be performed prior to entering the period of extended operation to determine the corrosivity of the soil in the vicinity ofnon-cathodically protected A.2.1.22 Within ten years prior to the steel pipe within the scope of this program. If the initial period of extended operation.

analysis shows the soil to be non-corrosive, this analysis will be re-performed every ten years thereafter.

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure 3 I Page 14

65.

Flux Thimble Tube

66.

Alkali-Silica Reaction (ASR)

Monitoring Program

67.

Structures Monitoring Program

68.

Structures Monitoring Program Implement measures to ensure that the movable incore detectors are not returned to service during the period of extended operation.

NextEra will perform an integrated review of expansion trends at Seabrook Station by conducting a periodic assessment of ASR expansion behavior to confirm that the MPR/FSEL large-scale test programs remain applicable to plant structures. This review will include the following specific considerations:

Review of all cores removed to date for trends of any indications of mid-plane cracking.

Comparison of in-plane expansion to through-thickness expansion of all monitored points by plotting these data on a graph of CCI versus through-thickness expansion.

Comparison of in-plane expansions and through-thickness expansions recorded to date to the limits from the MPR/FSEL large-scale test programs and check of margin for future expansion. Also, the calculated volumetric expansion will be compared to the range observed in the beam test programs and margin for future expansion will be checked.

Perform one shallow core bore in an area that was continuously wetted from borated water to be examined for concrete degradation and also expose rebar to detect any degradation such as loss of material. The removed core will also be subjected to petrographic examination for concrete degradation due to ASR per ASTM Standard Practice C856.

Perform sampling at the leak off collection points for chlorides, sulfates, pH and iron once every three months.

Prior to the period of extended operation. - In NIA Progress At least 5 years prior to the A.2.1.31.A period of extended operation and every 10 years thereafter.

A.2.1.31 Complete A.2.1.31 Complete

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure 3 I Page 15 Open-Cycle Cooling Water

69.

System Closed-Cycle Cooling Water

70.

System Alkali-Silica Reaction (ASR)

71.

Monitoring Program I Building Deformation Monitoring Program

72.

Flow-Accelerated Corrosion Inspection oflnternal Surfaces in

73.

Miscellaneous Piping and Ducting Components Replace the Diesel Generator Heat Exchanger Plastisol PVC lined Service Water piping with piping fabricated A.2.1.11 Complete from AL6XN material.

Inspect the piping downstream of CC-V-444 and CC-V-446 to determine whether the loss of material due to Within ten years prior to the cavitation induced erosion has been eliminated or whether A.2.1.12 period of extended operation.

this remains an issue in the primary component cooling water system.

NextEra has completed testing at the University of Texas Ferguson Structural Engineering Laboratory which demonstrates the parameters being monitored and acceptance criteria used are appropriate to manage the effects of ASR.

A.2.l.31A Prior to the period of NextEra will Implement the Alkali-Silica Reaction (ASR)

A.2.l.31B extended operation.

Monitoring Program and Building Deformation Monitoring Program described in B.2.1.3 lA and B.2.1.3 IB of the License Renewal Application.

Enhance the program to include management of wall A.2.1.8 Prior to the period of thinning caused by mechanisms other than F AC.

extended operation.

Enhance the program to include performance of focused examinations to provide a representative sample of20%,

or a maximum of25, of each identified material, A.2.1.25 Prior to the period of environment, and aging effect combinations during each extended operation.

I 0 year period in the period of extended operation.

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure 3 I Page 16

74.

Fire Water System

75.

Fire Water System

76.

Fire Water System Enhance the program to perform sprinkler inspections annually per the guidance provided in NFPA 25 (2011 Edition). Inspection will ensure that sprinklers are free of corrosion, foreign materials, paint, and physical damage and installed in the proper orientation (e.g., upright, pendant, or sidewall). Any sprinkler that is painted, corroded, damaged, loaded, or in the improper orientation, and any glass bulb sprinkler where the bulb has emptied, will be evaluated for replacement.

Enhance the program to a) conduct an inspection of piping and branch line conditions every 5 years by opening a flushing connection at the end of one main and by removing a sprinkler toward the end of one branch line for the purpose of inspecting for the presence of foreign organic and inorganic material per the guidance provided in NFPA 25 (2011 Edition) and b) If the presence of sufficient foreign organic or inorganic material to obstruct pipe or sprinklers is detected during pipe inspections, the material will be removed and its source is determined and corrected.

In buildings having multiple wet pipe systems, every other system shall have an internal inspection of piping every 5 years as described in NFPA 25 (2011 Edition), Section 14.2.2.

Enhance the Program to conduct the following activities annually per the guidance provided in NFP A 25 (2011 Edition).

main drain tests deluge valve trip tests fire water storage tank exterior surface inspections Prior to the period of A.2.1.16 extended operation.

Prior to the period of A.2.1.16 extended operation.

A.2.1.16 Prior to the period of extended operation.

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure 3 I Page 17

77.

Fire Water System

78.

External Surfaces Monitoring Open-Cycle Cooling Water

79.

System

80.

Fire Water System

81.

Fuel Oil Chemistry The Fire Water System Program will be enhanced to include the following requirements related to the main drain testing per the guidance provided in NFPA 25 (2011 Edition).

The requirement that if there is a 10 percent reduction Prior to the period of in full flow pressure when compared to the original A.2.1.16 extended operation.

acceptance tests or previously performed tests, the cause of the reduction shall be identified and corrected if necessary.

Recording the time taken for the supply water pressure to return to the original static (nonflowing) pressure.

Enhance the program to include periodic inspections of in-scope insulated components for possible corrosion under insulation. A sample of outdoor component surfaces that are insulated and a sample of indoor insulated components A.2.1.24 Prior to the period of exposed to condensation (due to the in-scope component extended operation.

being operated below the dew point), will be periodically inspected every 10 years during the period of extended operation.

Enhance the program to include visual inspection of Within 10 years prior to the internal coatings/linings for loss of coating integrity.

A.2.1.11 period of extended operation.

Enhance the program to include visual inspection of Within 10 years prior to the internal coatings/linings for loss of coating integrity.

A.2.1.16 period of extended operation.

Enhance the program to include visual inspection of Within 10 years prior to the internal coatings/linings for loss of coating integrity.

A.2.1.18 period of extended operation.

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure 3 I Page 18 Inspection oflntemal Surfaces in

82.

Miscellaneous Piping and Ducting Components

83.

Alkali-Silica Reaction Monitoring ASME Section XI, Subsection

84.

IWL

85.

Fire Water System

86.

Fire Water System

87.

Fire Water System Enhance the program to include visual inspection of Within I 0 years prior to the internal coatings/linings for loss of coating integrity.

A.2.1.25 period of extended operation.

Enhance the ASR AMP to install extensometers in all Tier 3 areas of two dimensional reinforced structures to monitor expansion due to alkali-silica reaction in the out-of-plane direction.

Monitoring expansion in the out-of-plane direction will A.2.l.31A February 28, 2017 commence upon installation of the extensometers and continue on a six month frequency through the period of extended operation.

Evaluate the acceptability of inaccessible areas for Prior to the period of structures within the scope of ASME Section XI, A.2.1.28 Subsection IWL Program.

extended operation.

Enhance the program to perform additional tests and inspections on the Fire Water Storage Tanks as specified in Section 9.2.7 ofNFPA 25 (2011 Edition) in the event Prior to the period of that it is required by Section 9.2.6.4, which states "Steel A.2.1.16 tanks exhibiting signs of interior pitting, corrosion, or extended operation.

failure of coating shall be tested in accordance with 9.2.7."

Enhance the program to include disassembly, inspection, A.2.1.16 Prior to the period of and cleaning of the mainline strainers every 5 years.

extended operation.

Increase the frequency of the Open Head Spray Nozzle Air Flow Test from every 3 years to every refueling A.2.1.16 Prior to the period of outage to be consistent with LR-ISG-2012-02, AMP extended operation.

XI.M27, Table 4a.

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure 3 I Page 19

88.

Fire Water System Inspection of Internal Surfaces in

89.

Miscellaneous Piping and Ducting Components PWR Vessel Internals

90.

Enhance the program to include verification that a) the drain holes associated with the transformer deluge system are draining to ensure complete drainage of the system after each test, b) the deluge system drains and associated piping are configured to completely drain the piping, and c) normally-dry piping that could have been wetted by inadvertent system actuations or those that occur after a fire are restored to a dry state as part of the suppression system restoration.

Incorporate Coating Service Level III requirements into the RCP Motor Refurbishment Specification for the internal painting of the motor upper bearing coolers and motor air coolers. All four RCP motors will be refurbished and replaced using the Coating Service Level III requirements prior to entering the period of extended operation.

Implement the PWR Vessel Internals Program. The program will be implemented in accordance with MRP-227-A (Pressurized Water Reactor Internals Inspection and Evaluation Guidelines) and NEI 03-08 (Guideline for the Management of Materials Issues).

Within five years prior to the A.2.1.16 period of extended operation.

Prior to the period of A.2.1.25 extended operation.

Prior to the period of A.2.1.7 extended operation

U.S. Nuclear Regulatory Commission SBK-L-17133 I Enclosure 3 I Page 20 91 Building Deformation Monitoring Implement the Building Deformation Monitoring Program Enhance the Structures Monitoring Program to require structural evaluations be performed on buildings and components affected by deformation as necessary to ensure that the structural function is maintained. Evaluations of structures will validate structural performance against the design basis, and may use results from the large-scale test programs, as appropriate. Evaluations for structural deformation will also consider the impact to functionality of affected systems and components (e.g., conduit expansion joints). NextEra will evaluate the specific circumstances against the design basis of the affected system or component.

Enhance the Building Deformation AMP to include A.2.l.31B March 15, 2020 additional parameters to be monitored based on the results of the CEB Root Cause, Structural Evaluation and walk downs. Additional parameters monitored will include:

alignment of ducting, conduit, and piping; seal integrity; laser target measurements; key seismic gap measurements; and additional instrumentation.

Develop a design standard to implement Aging Management Program B.2.1.31 B Building Deformation, Program Element 3 - Parameters Monitored/Inspected.

The design standard will clarify the deformation evaluation process and provide an auditable format to assess it. The design standard will include steps for each of the three evaluation stages that include parameters monitored, basis for why the parameter is monitored, and conditions that prompts action for the subsequent step.