SBK-L-12084, Response to Request for Additional Information NextEra Energy Seabrook License Renewal Application Supplemental Response - RAI B.2.1.11-2 and B.2.1.12-6

Response to Request for Additional Information NextEra Energy Seabrook License Renewal Application Supplemental Response - RAI B.2.1.11-2 and B.2.1.12-6

ML121220298
Person / Time
Site:	Seabrook
Issue date:	04/26/2012
From:	Freeman P NextEra Energy Seabrook
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SBK-L-12084
Download: ML121220298 (25)

Text

NExTera ENERGY_

SEABROOK April 26, 2012 SBK-L- 12084 Docket No. 50-443 U.S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852 Seabrook Station Response to Request for Additional Information NextEra Energy Seabrook License Renewal Application Supplemental Response - RAI B.2.1.11-2 and B.2.1.12-6

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- 11002, "Seabrook Station Response to Request for Additional Information, NextEra Energy Seabrook License Renewal Application Aging Management Programs - Set 4", January 13, 2011. (Accession Number ML110140809)

3. NextEra Energy Seabrook, LLC letter SBK-L-12023, "Seabrook Station Response to Request for Additional Information, NextEra Energy Seabrook License Renewal Application - Set 18", February 7, 2012 (Accession Number ML12041A054)

4. NextEra Energy Seabrook, LLC letter SBK-L-11063, "Seabrook Station Response to Request for Additional Information, NextEra Energy Seabrook License Renewal Application - Set 13", April 14, 2011. (Accession Number ML 1108A131)

5. NextEra Energy Seabrook, LLC letter SBK-L-11003, "Seabrook Station Response to Request for Additional Information, NextEra Energy Seabrook License Renewal Application Aging Management Programs - Set 5 ", January 13, 2011. (Accession Number (ML110140587)

6. NextEra Energy Seabrook, LLC letter SBK-L- 11027, "Seabrook Station Response to Request for Additional Information, NextEra Energy Seabrook License Renewal Application - Set 9", February 18, 2011. (Accession Number ML110530481)

K1 NextEra Energy Seabrook, LLC, P.O. Box 300, Lafayette Road, Seabrook, NH 03874

United States Nuclear Regulatory Commission SBK-L-12084 / Page 2 In Reference 1, NextEra Energy Seabrook, LLC (NextEra) submitted an application for a renewed facility operating license for Seabrook Station Unit 1 in accordance with the Code of Federal Regulations, Title 10, Parts 50, 51, and 54.

In Reference 2, NextEra provided a response to RAI B.2.1.12-6 related to the Closed-Cycle Cooling Water System Program. In Reference 3, NextEra provided a response to request for additional information (RAI) B.2.1.11-2 related to the Open-Cycle Cooling Water System Program. Based on questions from the staff in a teleconference on April 10, 2012, Enclosed are enhanced responses to RAIs B.2.1.11-2 and B.2.1.12-6. Commitment numbers 69 and 70 are added to the License Renewal Commitment List by this letter. For clarity, deleted text is highlighted by strikethrough and inserted text is highlighted in bold italic.

In Reference 4, the schedule was changed for implementation of commitment number 52, "Implement measures to maintain the exterior surface of the Containment Structure, from elevation -30feet to +20feet dewatered". The revised schedule date, "By December 31, 2012" was inadvertently changed in subsequent LRA Table A.3 transmittals. The LRA Table A.3 has been revised to reflect the correct schedule date of commitment 52. A review of NextEra correspondence to date was performed and two minor editorial differences (commitments 60, Reference 5 and 62, Reference 6) were found and LRA Table A.3 has been corrected. provides the current LRA Appendix A - Final Safety Report Supplement Table A.3, License Renewal Commitment List, updated to reflect the license renewal commitment changes made in NextEra Energy Seabrook correspondence to date.

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

Cliche, License Renewal Project Manager, at (603) 773-7003.

If you have any questions regarding this correspondence, please contact Mr. Michael O'Keefe, Licensing Manager, at (603) 773-7745.

Sincerely, NextEra Energy Seabrook, LLC.

Paul 0. Freeman Site Vice President

Enclosures:

- Supplemental Response to Requests for Additional Information B.2.1.11-2 and B.2.1.12-6 Seabrook Station License Renewal Application - LRA Appendix A - Final Safety Report Supplement Table A.3, License Renewal Commitment List, updated to reflect the license renewal commitment changes made in NextEra Seabrook correspondence to date.

United States Nuclear Regulatory Commission SBK-L-12084/Page 3 cc:

W.M. Dean, NRC Region I Administrator J. G. Lamb, NRC Project Manager, Project Directorate 1-2 W. J. Raymond, NRC Resident Inspector A.D. Cunanan, NRC Project Manager, License Renewal M. Wentzel, NRC Project Manager, License Renewal Mr. Christopher M. Pope 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 John Giarrusso, Jr., Nuclear Preparedness Manager The Commonwealth of Massachusetts Emergency Management Agency 400 Worcester Road Framingham, MA 01702-5399

United States Nuclear Regulatory Commission SBK-L-12084/ Page 4 N~x~era" ENERG-2-AQ/ SEABROOK I, Paul 0. Freeman, Site Vice President of NextEra Energy Seabrook, LLC hereby affirm that the information and statements contained within are based on facts and circumstances which are true and accurate to the best of my knowledge and belief.

Sworn and Subscribed Before me this day of April, 2012 Paul 0. Freeman Site Vice President Notary Puilic /

Enclosure 1 to SBK-L-12084 Supplemental Response to Requests for Additional Information B.2.1.11-2 and B.2.1.12-6 Seabrook Station License Renewal Application

United States Nuclear Regulatory Commission Page 2 of 8 SBK-L-12084 / Enclosure 1 Request for Additional Information (RAI) B.2.1.11-2

Background

GALL Report AMP XI.M20, "Open-Cycle Cooling Water," states that the program includes surveillance and control techniques to manage aging effects caused by various aging mechanisms including protective coating failures. GALL Report, Table IX.F, "Aging Mechanisms," states that fouling includes macrofouling (e.g., peeled coatings and debris), and can result in a reduction of heat transfer or loss of material.

SRP-LR Section A.1.2.3.10 "Operating Experience," states that past corrective actions for existing AMPs should be considered and that feedback from past failures should have resulted in appropriate program enhancements. The SRP-LR also states that operating experience information should provide objective evidence to support the conclusion that the effects of aging will be managed adequately so that the structure and component intended function(s) will be maintained during the period of extended operation.

LRA Section B.2.1.11 describes the Open-Cycle Cooling Water System Program as an existing program that manages the aging effects due to various mechanisms including "liner/coating degradation." In addition, the Operating Experience section for this program states:

The cement lined above groundpiping associatedwith the Diesel Generatorheat exchangers has been replaced with flanged Plastisol PVC lined carbon steel spool pieces. The size and accessibility of this piping did not permit the use of AMEX-I0/WEKO seals. Follow up inspections of weld areas by ultrasonic testing and internal visual examinations during refueling outages have confirmed that the engineering design change has been effective in preventing loss of material.

Issue According to recent information provided by Regional NRC personnel, the Plastisol PVC lining has degraded to the extent that it was found missing in certain portions of the carbon steel piping, which potentially affected the intended function of the diesel generator heat exchangers. Based on this plant-specific operating experience, additional information is needed by the staff relative to the effectiveness of past aging management activities for the Open-Cycle Cooling Water System Program, and any enhancements, if warranted, to address this degradation.

Request (a) Provide a description of the recent PVC lining degradation event, including the associated cause and extent of condition. As part of the response, address the expected life span of the PVC lining material. In addition, provide a discussion of the previous aging management activities that were performed to manage liner degradation prior to the event, including whether any previous activities were specifically performed on the degraded areas.

(b) Provide a description of the corrective actions taken in response to the recent event and provide any enhancements made to the Open-Cycle Cooling Water Program to ensure that

United States Nuclear Regulatory Commission Page 3 of 8 SBK-L-12084 / Enclosure 1 components' intended function(s) will not be impacted during the period of extended operation. If enhancements will not be made, provide the bases for why there is reasonable assurance that the intended functions will be maintained consistent with current licensing bases.

NextEra Energy Response to Request (a)

Background

During Refueling Outage 2 (1992), several through wall leaks were observed in the 16" cement lined Service Water piping for the Diesel Generator heat exchangers (DGHXs). Subsequently, in Refueling Outage 3 (1994), the cement lined carbon steel Service Water piping associated with the Diesel Generator heat exchangers was replaced with Plastisol PVC lined carbon steel piping.

At the time that the design change was initiated, the liner material was noted to have an anticipated service life of 15 years. Subsequent discussion with the liner manufacturer indicated an expected life of 15 - 20 years for the Plastisol PVC liner in seawater.

During routine surveillance testing in July 2011, Service Water flow through the Train 'B' DGHX was identified as degraded. At that time, the apparent cause of the flow degradation was postulated to be macrofouling of the heat exchanger or flow orifice. A corrective action document was initiated to inspect the heat exchanger and flow orifice during the next outage. In October 2011, the plant entered a forced outage for unrelated reasons and the Train 'B' DGHX downstream flow orifice was inspected. Inspection revealed that pieces of Plastisol PVC lining of sufficient size so as to partially restrict flow through the orifice had become detached from the pipe.

Description of PVC Liner Degradation Event Inspection of the Plastisol PVC lined piping revealed the majority of the lining in one Service Water spool piece downstream of the Train 'B' DGHX to be in poor condition. The lining was loose or missing in multiple locations. This spool piece had been previously inspected during Refueling Outage 6 (1999) with no deficiencies noted. All remaining liner material installed in this spool piece was removed to prevent further delamination.

The remainder of the Plastisol PVC lined pipe in the Train 'B' DGHX piping (with the exception noted below) and all of the Train 'A' DGHX piping was inspected and the results did not identify similar delamination. One section of 'B' DGHX pipe was not inspected based on acceptable UT measurements and boroscope inspection results associated with the adjacent vertical section of piping. Localized areas were noted to show bubbles or small waves in the liner. The Plastisol PVC liner in these areas was removed and edges smoothed.

In lieu of repair, the Plastisol PVC lined piping has been scheduled for replacement with a corrosion resistant, unlined material during the next refueling outage. Evaluation of the existing unlined portions of carbon steel pipe as a result of liner removal indicates sufficient wall thickness for continued operation until the piping can be replaced.

Associated Cause A sample of the liner was sent to a materials analysis laboratory for examination. Results of that examination concluded that the Plastisol PVC liner material removed from the Service Water system exhibited indications of aging. This was evidenced by cracking on both the OD (pipe wall side) and ID (seawater side) of the removed liner material. Hardness testing was performed on

United States Nuclear Regulatory Commission Page 4 of 8 SBK-L-12084 / Enclosure 1 the section of liner material removed from the Service Water system and on an unused sample of liner material that was available from 1994 testing at the same laboratory. This test indicated a noticeable increase in hardness of the used liner material compared to the unused sample.

NextEra conducted a root cause evaluation of the Plastisol PVC lining degradation event. The root cause evaluation concluded that the lack of formal process to track the inspections and replacement strategy for the Plastisol PVC lined pipe and the failure of the inspection strategy to be consistently implemented over the past 17 years had led to the unexpected delamination of the DGHX Plastisol PVC lined pipe as discovered in October 2011.

This evaluation identified two root causes for this event.

1. A "limited life" design change (the Plastisol PVC material had a 15 year service life) was implemented in 1994 with no provisions for formally tracking the periodic verifications of the material condition of the coating.

2. Oversight of the Service Water system was not adequate due to a lack of compliance with the system performance monitoring guideline requirements associated with the Plastisol PVC lined pipe.

Previous Aging Management Activities and Inspections Following the piping replacement in 1994, the responsible System Engineer implemented an inspection strategy for this piping intended to ensure that a portion of the Plastisol PVC lined piping was inspected during each outage with each subsequent inspection scope and frequency being driven from previous inspection results. Inspection of the piping was performed each outage from Refueling Outage 4 (1996) though Refueling Outage 9 (2003) with no significant indications of liner degradation except as discussed below. Minor defects noted were determined to be caused by spool piece removal and reinstallation.

Results of piping inspection in Refueling Outage 5 (1997) noted some bubbles in the Plastisol PVC liner and delamination at the flange. An action was assigned to perform a more thorough inspection during the next outage (Refueling Outage 6) to determine if a generic problem existed for Plastisol PVC liner based on these inspection results.

The DGHX Train 'B' Plastisol PVC lined piping was inspected during Refueling Outage 6 (1999). Inspection results identified overall liner conditions to be "good" with no indications of deterioration. Two small areas of liner degradation at flange locations were identified and repaired. The System Engineer documented that the damage mechanism was suspected to be mechanical in nature and likely occurred during previous pipe spool removal/installation.

Areas of bubbling were identified but no significant indications of the Plastisol PVC liner separating from the pipe surface were observed. It was concluded that there is no generic problem existing for the Plastisol PVC liner.

DGHX Train 'B' Plastisol PVC lined piping was inspected during Refueling Outage 8 (2002). Inspection results identified damage to Plastisol PVC lined pipe upstream of an expansion joint. A portion of the Plastisol PVC liner in the individual spool piece was removed and repaired. Inspection notes stated that there was lack of adhesion of the liner to the pipe surface.

Following Refueling Outage 9 (2003), periodic inspection of the Plastisol PVC lined piping was discontinued in favor of a new long term inspection strategy. This strategy focused on the

United States Nuclear Regulatory Commission Page 5 of 8 SBK-L-12084 / Enclosure 1 Service Water system as a whole and distributed piping inspections over several outages. The Plastisol PVC lined pipe was not singled out for more frequent inspections. In accordance with this long term strategy, the DGHX piping (Plastisol PVC lined) was not scheduled for another inspection until Refueling Outages 15 and 16 (2012 and 2014, respectively).

NextEra EnerUy Response to Request (b)

Description of the Corrective Actions Taken as a Result of this Event Corrective Actions have been assigned to ensure that "limited life" design changes include adequate post-modification inspection and replacement activities to preclude similar unanticipated failures.

1. The design control process will-be has been revised to include the requirement to utilize the preventive maintenance process for inspections and replacement activities.

2. A pr.oess will be established that Plant Engineering Performance Monitoring Guidelines and System Turnover Guidelines have been revised to ensure enstures monitoring and inspection programs comply with system performance monitoring guidelines and that long term strategies comply with Regulatory Commitments (such as GL 89-13).

To resolve the issues in the DGHX Plastisol PVC lined piping, actions have been assigned to support replacement of the Plastisol PVC lined pipe in the subsequent refueling outage (currently scheduled for fall of 2012).

1. A design change document will be developed to resolve the degraded/nonconforming conditions.

2. The associated Service Water piping will be periodically inspected to verify adequate pipe wall thickness until replaced.

AL6XN will be utilized as the replacement material. The Service Water system already has AL6XN piping installed elsewhere in the system. In the LRA, AL6XN was treated as stainless steel and includes stainless steel piping and fittings in Condensation and Raw Water environments as shown on Table 3.3.2-37, pages 468 and 469 of the LRA. Therefore, no changes are required to the AMR line items. Replacement of both Trains of PlastisolPVC lined Service Waterpiping is currently scheduledfor the upcoming refueling outage in fall of 2012. The following commitment has been added to Section A.3 License Renewal Commitment List as follows:

NO. PROGRAM or COMMITMENT UFSAR TOPIC LOCATION Replace the Diesel Generator Priorto the Open-Cycle Heat Exchanger PlastisolPVC periodof 69 Cooling Water lined Service Water piping with A.2.1.11 extended System pipingfabricatedfrom AL6XN operation.

material To verify the extent of condition, actions have been assigned to evaluate concrete lined piping in the Screen Wash System and Circulating Water System for determination of liner adequacy and

United States Nuclear Regulatory Commission Page 6 of 8 SBK-L-12084 / Enclosure 1 to determine if other coatings utilized in the Service Water system have service life limitations. If any exist, corrective actions will be initiated to determine extent of condition and to develop a formal process to periodically verify the material condition of the coatings.

Belzonaproducts are polymeric materials commonly used for lining and liner repairs at Seabrook Station. An engineering evaluation performed in 1993 indicates that Belzona lined pipe has an expected service life of 15 years. However, review of the NextEra Energy OE databasehas not indicatedfailures of the Belzona lining due to exceeding its service life. The condition of all Service Water pipe linings is monitored via periodic internalpipe inspections in accordance with the "Service Water Inspection and Repair Trending Program"' Preventive maintenance activities have been initiatedto insure inspections are scheduled and are being tracked. The extent of use of the Belzona and polyurethane coatings in Service Water piping has been identified and is being tracked in the Service Water Inspection and Repair Trending Program.

Portions of Screen Wash System and Circulating Water System are in-scope for NSAS and have cement lining similar to the Service Water System. The internalsurfaces of the LR in-scope portions of Circulating Water System are managed under the Open-Cycle Cooling Water Programsas described on LRA Table 3.4.2-4, pages 3.4-67 through 3.4-69. The Screen Wash System is not within the scope of GL 89-13 and therefore, is not managed under the Open Cycle Cooling Water System. However, the internalsurfaces of the LR in-scopeportions of the Screen Wash System are managed under the Inspection of Internal Surfaces in Miscellaneous Pipingand Ducting Components Programas described in LRA Table 3.3.2-36, pages 3.3-457 through3.3-460.

An action has also been assigned to issue an Industry Operating Experience Report documenting this event.

Review of Open-Cycle Cooling Water Program to Determine if Enhancements are Needed NextEra Open-Cycle Cooling System Program (B.2.1.12) has provisions for managing protective coatings and therefore, enhancements are not required to the aging management program. The program relies on the implementation of the recommendations of NRC Generic Letter (GL) 89-13, "Service Water System Problems Affecting Safety-Related Equipment" to ensure that the aging effects on the open-cycle cooling water systems will be adequately managed for the period of extended operation. The program, as mandated by GL 89-13, includes (a) surveillance and control of corrosion, erosion, protective coating failure, bio-fouling, silting, and heat transfer degradation, (b) tests to verify heat transfer, (c) routine inspection and maintenance of plant components, (d) system walk downs to ensure compliance with the stations licensing basis and (e) a review of maintenance, operating and training practices and procedures to ensure the effectiveness of established programs.

The operating experience described above and the actions taken ensure that routine inspection and maintenance of this piping will be conducted appropriately for the applicable piping material, liner material, and design considerations. Replacing the Plastisol PVC lined piping (currently planned for Refueling Outage 15 in October 2012) prior to entering the Period of Extended Operation ensures that failure of this lining material does not become a viable aging mechanism requiring management.

United States Nuclear Regulatory Commission Page 7 of 8 SBK-L-12084 / Enclosure 1 Request for Additional Information (RAI) B.2.1.12-6

Background

The SRP-LR states that past operating experience would not necessarily invalidate an aging management program because the feedback from operating experience should have resulted in appropriate program enhancements or new programs. A review of past operating experience indicated a recurring condition in the primary component cooling water system with loss of material in piping downstream of valves CC-V-444 (CR 05-04881) and CC-V-446 (CR 03-01549) apparently due to cavitation erosion from throttling. The applicant stated that it had conducted flow rebalancing to alleviate the concern.

Issue It was not clear to the staff how the applicant has re-evaluated these areas after flow rebalancing was conducted to determine whether loss of material due to cavitation erosion remains an issue in the primary component cooling water system.

Request Provide additional information on how the loss of material due to cavitation erosion was confirmed to have been eliminated or whether this remains an issue in the primary component cooling water system. If loss of material for this mechanism is still an applicable aging issue, provide information on what program is managing this aging effect and how.

NextEra Energy Seabrook Response:

Flow re-balancing of the system performed in 2003 was expected to eliminate the cavitation-induced wear downstream of the throttled butterfly valves in question, and therefore, no follow-up inspections were scheduled as part of the corrective action process.

To ascertain whether the loss of material due to cavitation erosion has been eliminated or whether this remains an issue in the primary component cooling water system, the piping downstream of these valves will be inspected for loss of material prior to entering the period of extended operation. This inspection will be performed during the ten year period prior to the period of extended operation to allow adequate time for the condition to re-appear following the prior repairs.

A condition report has been initiated to perform these inspections during the ten year period prior to entering the period of extended operation. Additionally, the following commitment has been added to Section A.3 License Renewal Commitment List asfollows:

United States Nuclear Regulatory Commission Page 8 of 8 SBK-L-12084 / Enclosure 1 NO. PROGRAM or COMMITMENT UFSAR SCHEDULE TOPIC LOCATION Inspect the piping downstream of CC- V-444 and CC-V-446 to Within ten years determine whether the loss of priort te yhe Closed-Cycle materialdue to cavitation induced poth 70 Cooling Water erosion has been eliminated or A.2.1.12 period of System whether this remains an issue in extended the primary component cooling operation.

water system.

Enclosure 2 to SBK-L-12084 LRA Appendix A - Final Safety Report Supplement Table A.3 License Renewal Commitment List

United States Nuclear Regulatory Commission Page 2 of 13 SBK-L-12084 / Enclosure 2 A.3 LICENSE RENEWAL COMMITMENT LIST UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Program to be implemented prior to the period of extended operation.

Inspection plan to be submitted to NRC not later

1. PWR Vessel Internals An inspection plan for Reactor Vessel Internals will be A.2.1.7 than 2 years after receipt of submitted for NRC review and approval, the renewed license or not less than 24 months prior to the period of extended operation, whichever comes first.

Enhance the program to include visual inspection for Prior to the period of

2. Closed-Cycle Cooling cracking, loss of material and fouling when the in-scope A.2.1.12 Prirntotheperiod systems are opened for maintenance.

Inspection of Overhead Heavy Load and Light Enhance the program to monitor general corrosion on the Prior to the period of

3. Load (Related to crane and trolley structural components and the effects of A.2.1.13 extended operation Refueling) Handling wear on the rails in the rail system.

Systems Inspection of Overhead Heavy Load and Light Enhance the program to list additional cranes for Prior to the period of

4. Load (Related to monitoring. A.2.1.13 extended operation Refueling) Handling Systems Compressed Air Enhance the program to include an annual air quality test Prior to the period of

5. Monitoring requirement for the Diesel Generator compressed air sub A.2.1.14 extended operation system.

United States Nuclear Regulatory Commission Page 3 of 13 SBK-L-12084 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION

6. Fire Protection Enhance the program to perform visual inspection of A.2.1.15 Prior to the period of penetration seals by a fire protection qualified inspector, extended operation.

Enhance the program to add inspection requirements such

7. Fire Protection as spalling, and loss of material caused by freeze-thaw, A2115 Prior to the period of chemical attack, and reaction with aggregates by qualified extended operation.

inspector.

Enhance the program to include the performance of visual Prior to the period of

8. Fire Protection inspection of fire-rated doors by a fire protection qualified A.2.1.15 extended operation.

inspector.

Enhance the program to include NFPA 25 guidance for

'where sprinklers have been in place for 50 years, they Prior to the period of

9. Fire Water System shall be replaced or representative samples from one or A.2.1.16 extended operation.

more sample areas shall be submitted to a recognized testing laboratory for field service testing".

Enhance the program to include the performance of Prior to the period of

10. Fire Water System periodic flow testing of the fire water system in accordance A.2.1.16 extended operation.

with the guidance of NFPA 25.

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. These inspections will be documented and trended to determine if

11. Fire Water System a representative number of inspections have been A.2.1.16 Within ten years prior to the performed prior to the period of extended operation. If a period of extended operation.

representative number of inspections have not been performed prior to the period of extended operation, focused inspections will be conducted. These inspections will be performed within ten years prior to the period of extended operation.

United States Nuclear Regulatory Commission Page 4 of 13 SBK-L-12084 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Enhance the program to include components and aging

12. Aboveground Steel effects required by the Aboveground Steel Tanks. A.2.1.17 Prior to the period of Tanks extended operation.

13. Aboveground Steel Enhance the program to include an ultrasonic inspection and evaluation of the internal bottom surface of the two Fire A.2.1.17 Within ten years prior to the Tanks Protection Water Storage Tanks. period of extended operation.

Enhance program to add requirements to 1) sample and analyze new fuel deliveries for biodiesel prior to offloading Prior to the period of

14. Fuel Oil Chemistry to the Auxiliary Boiler fuel oil storage tank and 2) A.2.1.18 extended operation.

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

Enhance the program to add requirements to check for the

15. Fuel Oil Chemistry presence of water in the Auxiliary Boiler fuel oil storage A.2.1.18 Prior to the period of tank at least once per quarter and to remove water as extended operation.

necessary.

Enhance the program to require draining, cleaning and Prior to the period of

16. Fuel Oil Chemistry inspection of the diesel fire pump fuel oil day tanks on a A.2.1.18 extended operation.

frequency of at least once every ten years.

Enhance the program to require ultrasonic thickness measurement of the tank bottom during the 10-year

17. Fuel Oil Chemistry draining, cleaning and inspection of the Diesel Generator A.2.1.18 Prior to the period of fuel oil storage tanks, Diesel Generator fuel oil day tanks, extended operation.

diesel fire pump fuel oil day tanks and auxiliary boiler fuel oil storage tank.

Reactor Vessel Enhance the program to specify that all pulled and tested Prior to the period of

18. SuRveillance capsules, unless discarded before August 31, 2000, are A.2.1.19 extended operation.

placed in storage.

United States Nuclear Regulatory Commission Page 5 of 13 SBK-L-12084 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Enhance the program to specify that if plant operations exceed the limitations or bounds defined by the Reactor Reactor Vessel Vessel Surveillance Program, such as operating at a lower Prior to the period of

19. Surveillance cold leg temperature or higher fluence, the impact of plant A.2.1.19 extended operation.

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 Reactor Vessel outage in which the capsule receives a neutron fluence that Prior to the period of

20. Surveillance meets the schedule requirements of 10 CFR 50 Appendix A.2.1.19 extended operation.

H and ASTM E185-82 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,

21. Reactor Vessel without the intent to test it, is stored in a manner which A.2.1.19 Prior to the period of Surveillance maintains it in a condition which would permit its future use, extended operation.

including during the period of extended operation.

Within ten years prior to the Implement the One Time Inspection Program. A.2.1.20 perio tended prio n.

22. One-Time Inspection period of extended operation.

Implement the Selective Leaching of Materials Program.

23.

Selective Leaching of The program will include a one-time inspection of selected Within five years prior to the Materials

23. components where selective leaching has not been A.2.1.21 period of extended operation.

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

24. Buried Piping And Tanks Program. A.2.1.22 entering the period of Inspection extended operation

United States Nuclear Regulatory Commission Page 6 of 13 SBK-L-12084 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION One-Time Inspection of Implement the One-Time Inspection of ASME Code Class Within ten years prior to the

25. ASME Code Class 1 A212 Sma Bore-Piping1 1 Small Bore-Piping Program. period of extended operation.

Small Bore-Piping Enhance the program to specifically address the scope of the program, relevant degradation mechanisms and effects External Surfaces of interest, the refueling outage inspection frequency, the Prior to the period of

26. Monitoring inspections of opportunity for possible corrosion under A.2.1.24 extended operation.

insulation, the training requirements for inspectors and the required periodic reviews to determine program effectiveness.

Inspection of Internal Surfaces in Surface s Piping Implement the Inspection of Internal Surfaces in Prior to the period of

27. Miscellaneous Miscellaneous Piping and Ducting Components Program. A.2.1.25 extended operation.

and Ducting Components Enhance the program to add required equipment, lube oil Prior to the period of

28. Lubricating Oil Analysis analysis required, sampling frequency, and periodic oil A.2.1.26 extended operation.

changes.

Enhance the program to sample the oil for the Reactor A.2.1.26 Prior to the period of

29. Lubricating Oil Analysis Coolant pump oil collection tanks. extended operation.

Enhance the program to require the performance of a one-

30. Lubricating Oil Analysis time ultrasonic thickness measurement of the lower portion A.2.1.26 Prior to the period of of the Reactor Coolant pump oil collection tanks prior to the extended operation.

period of extended operation.

31. ASME Section XI, Enhance procedure to include the definition of A.2.1.28 Prior to the period of Subsection IWL "Responsible Engineer". extended operation.

Structures Monitoring Enhance procedure to add the aging effects, additional Prior to the period of

32. Program locations, inspection frequency and ultrasonic test A.2.1.31 extended operation.

requirements.

United States Nuclear Regulatory Commission Page 7 of 13 SBK-L-12084 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Structures Monitoring Enhance procedure to include inspection of opportunity Prior to the period of

33. Program when planning excavation work that would expose A.2.1.31 extended operation.

inaccessible concrete.

Electrical Cables and Connections Not Subject Implement the Electrical Cables and Connections Not

34. 50.49 Subject tPrioro 10 CFR 50.49 Environmental Qualification A.2.1.32of Environmental Subjeto 1 ro50. extended operation.

Qualification Requirements program.

Requirements Electrical Cables and Connections Not Subject to 10 CFR 50.49 Implement the Electrical Cables and Connections Not Prior to the period of

35. Environmental Subject to 10 CFR 50.49 Environmental Qualification A.2.1.33 Prirntotheperiodo Qualification Requirements Used in Instrumentation Circuits program.

Requirements Used in Instrumentation Circuits Inaccessible Power Cables Not Subject to Implement the Inaccessible Power Cables Not Subject to

36. 10 CFR 50.49 10 CFR 50.49 Environmental Qualification Requirements A.2.1.34 Prior to the period of Environmental 1ro 50. extended operation.

Qualification program.

Requirements Prior to the period of

37. Metal Enclosed Bus Implement the Metal Enclosed Bus program. A.2.1.35 exte e perion.

extended operation.

Prior to the period of

38. Fuse Holders Implement the Fuse Holders program. A.2.1.36 exte e perion.

extended operation.

United States Nuclear Regulatory Commission Page 8 of 13 SBK-L-12084 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Electrical Cable Connections Not Subject Implement the Electrical Cable Connections Not Subject to Prior to the period of

39. to 10 CFR 50.49 10 CFR 50.49 Environmental Qualification Requirements A.2.1.37 exte e perion.

Environmental extended operation.

Qualification program.

Requirements Prior to the period of 345 KV SF 6 Bus Implement the 345 KV SF 6 Bus program. A.2.2.1 exte e perion.

40. extended operation.

Metal Fatigue of Reactor Enhance the program to include additional transients Prior to the period of

41. Coolant Pressure beyond those defined in the Technical Specifications and A.2.3.1 extended operation.

Boundary UFSAR.

Metal Fatigue of Reactor Enhance the program to implement a software program, to Prior to the period of

42. Coolant Pressure count transients to monitor cumulative usage on selected A.2.3.1 extended operation.

Boundary components.

Pressure -Temperature The updated analyses will be Limits, including Low Seabrook Station will submit updates to the P-T curves and submitted at the appropriate

43. Temperature LTOP limits to the NRC at the appropriate time to comply A.2.4.1.4 time to comply with 10 CFR Overpressure Protection with 10 CFR 50 Appendix G. 50 Appendix G, Fracture Limits Toughness Requirements.

United States Nuclear Regulatory Commission Page 9 of 13 SBK-L-12084 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION 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 of NUREG/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 Environmentally- water environment. This includes applying the appropriate At least two years prior to

44. Assisted Fatigue Fen factors to valid CUFs determined from an existing A.2.4.2.3 entering the period of Analyses (TLAA) fatigue analysis valid for the period of extended operation 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, if CUF, 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 the NRC).

United States Nuclear Regulatory Commission Page 10 of 13 SBK-L-12084 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION

45. Number Not Used Protective Coating Enhance the program by designating and qualifying an Prior to the period of

46. Monitoring and Inspector Coordinator and an Inspection Results Evaluator. A.2.1.38 extended operation Maintenance Enhance the program by including, "Instruments and Protective Coating Coan g Equipment needed for limited to, flashlight, inspection spotlights, may include, marker but not be pen, mirror, A.2.1.38 Prior to the period of

47. Monitoring and measuring tape, magnifier, binoculars, camera with or extended operation without wide angle lens, and self sealing polyethylene sample bags."

Protective Coating Enhance the program to include a review of the previous Prior to the period of

48. Monitoring and two monitoring reports. A.2.1.38 extended operation Maintenance Protective Coating Enhance the program to require that the inspection report

49. Monitoring and is to be evaluated by the responsible evaluation personnel, A.2.1.38 Prior to the period of Montonand who is to prepare a summary of findings and extended operation recommendations for future surveillance or repair.

Within the next two refueling outages, OR15 or OR16, and ASME Section XI, Perform UT testing of the containment liner plate in the repeated at intervals of no

50. Subsection IWE vicinity of the moisture barrier for loss of material. A.2.1.27 more than five refueling outages Number Not Used 51.

United States Nuclear Regulatory Commission Page 11 of 13 SBK-L-12084 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION ASME Section XI, Implement measures to maintain the exterior surface of the

52. Subsection IWL Containment Structure, from elevation -30 feet to +20 feet, A.2.1.28 By-2012 By December 31, in a dewatered state.

Replace the spare reactor head closure stud(s) Prior to the period of Reactor Head Closure manufactured from the bar that has a yield strength > 150 A.2.1.3 ex te e perion.

Studs

53. ksi with ones that do not exceed 150 ksi. extended operation.

Unless an alternate repair criteria changing the ASME code boundary is permanently approved by the NRC, or the Seabrook Station steam generators are changed to Program to be submitted to

54. Steam Generator Tube eliminate PWSCC-susceptible tube-to-tubesheet welds, A.2.1.10 NRC at least 24 months prior Integrity submit a plant-specific aging management program to to the period of extended manage the potential aging effect of cracking due to operation.

PWSCC at least twenty-four months prior to entering the Period of Extended Operation.

Steam Generator Tube Seabrook will perform an inspection of each steam Prior to entering the period of

55. Integrity generator to assess the condition of the divider plate A.2. 1.10 extended operation assembly.

56.

Closed-Cycle Cooling Revise the station program documents to reflect the EPRI Prior to entering the period of

56. Closed-CylemCooling Guideline operating ranges and Action Level values for A.2.1.12 Pitenteritio do Water System hydrazine and sulfates. extended operation.

Closed-Cycle Cooling Revise the station program documents to reflect the EPRI Prior to entering the period of

57. Water System Guideline operating ranges and Action Level values for A.2.1.12 extended operation.

Diesel Generator Cooling Water Jacket pH.

United States Nuclear Regulatory Commission Page 12 of 13 SBK-L-12084 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Update Technical Requirement Program 5.1, (Diesel Fuel Prior to the period of

58. Fuel Oil Chemistry Oil Testing Program) ASTM standards to ASTM D2709-96 A.2.1.18 extended operation.

and ASTM D4057-95 required by the GALL XI.M30 Rev 1 Nickel Alloy Nozzles and The Nickel Alloy Aging Nozzles and Penetrations program Prior to the period of Penetrations will implement applicable Bulletins, Generic Letters, and A.2.2.3 extended operation.

staff accepted industry guidelines.

Buried Piping and Tanks Implement the design change replacing the buried Auxiliary Prior to entering the period of

60. Inspection Boiler supply piping with a pipe-within-pipe configuration A.2.1.22 extended operation.

with leak 4nd*Gat~en detection capability.

Replace the flexible hoses associated with the Diesel Within ten years prior to

61. Compressed Air Generator air compressors on a frequency of every 10 A.2.1.14 entering the period of Monitoring Program years. extended operation.

Enhance the program to include a statement that sampling Prior to ente~r-g the period of

62. Water Chemistry frequencies are increased when chemistry action levels are A.2.1.2 extended operation.

exceeded.

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 CVCS Prior to the period of Charging Pump mini flow above the acceptance criteria 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 of non-cathodically protected steel pipe A.2.1.22 Prior to entering the period of

64. Buried Piping and Tanks within the scope of this program. If the initial analysis extended operation.

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

United States Nuclear Regulatory Commission Page 13 of 13 SBK-L-12084 / Enclosure 2 UFSAR No, PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Implement measures to ensure that the movable incore Prior to entering the period of

65. detectors are not returned to service during the period of Flux Thimble Tube extended operation. N/A extended operation Number Not Used 66.

Perform one shallow core bore in an area that was

67. Structures Monitoring continuously wetted from borated water to be examined for A.2.1.31 No later than December 31, Program concrete degradation and also expose rebar to detect any 2015 degradation such as loss of material.

Perform sampling at the leakoff collection points for

68. Structures Monitoring Program chlorides, sulfates, pH and iron once every three months. A.2.1.31 Starting January 2014 Replace the Diesel GeneratorHeat Exchanger Prior to the period of

69. Open-Cycle Cooling PlastisolPVC lined Service Water piping with piping A.2.1.11 extended operation.

Water System fabricatedfrom AL6XN material.

Inspect the piping downstream of CC-V-444 and CC-Closed-Cycle Cooling V-446 to determine whether the loss of material due Within ten years prior to

70. Water System to cavitation induced erosion has been eliminated or A.2.1.12 the period of extended whether this remains an issue in the primary operation.

component cooling water system.