SBK-L-14089, Response to Request for Additional Information Related to the Review of the Seabrook Station License Renewal Application- Set 21 (Tag ME4028)

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Response to Request for Additional Information Related to the Review of the Seabrook Station License Renewal Application- Set 21 (Tag ME4028)
ML14177A502
Person / Time
Site: Seabrook NextEra Energy icon.png
Issue date: 06/24/2014
From: Dean Curtland
NextEra Energy Seabrook
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
SBK-L-14089, TAC ME4028
Download: ML14177A502 (61)
v • d • e


Text

NExTerao ENERGYj Q4 June 24, 2014 10 CFR 54 SBK-L-14089 Docket No. 50-443 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Seabrook Station Response to Request for Additional Information Related to the Review of The Seabrook Station License Renewal Application- Set 21 (Tag No. ME4028)

References:

1. NextEra Energy Seabrook letter SBK-L-10077, "Seabrook Station Application for Renewed Operating License", May 25, 2010. (Accession Number ML101590099).
2. EPRI Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev. 0), Technical Report 1016596, December, 2008.
3. NRC Regulatory Issue Summary 2011-07, License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management, July 21, 2011.
4. Revision 0 of the Safety Evaluation Report for EPRI Materials Reliability Program:

Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev. 0), June 22, 2011.

5. EPRI Materials Reliability Program:.Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A), Technical Report 1022863, December, 2011.
6. Revision 1 to the Safety Evaluation Report for EPRI Materials Reliability Program:

Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A), December 16, 2011.

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

U.S. Nuclear Regulatory Commission SBK-L-14089/Page 2

7. LR-ISG-2011-04, "Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors", May 28, 2013.
8. NRC Letter, Requests For Additional Information Related to the Review of the Seabrook Station, License Renewal Application- Set 21 (TAC NO. ME4028), April 25, 2014, (Accession Number ML14101A324).
9. NextEra Energy Seabrook letter SBK-L- 11063, "Response to Request for Additional Information, NextEra Energy Seabrook License Renewal Application, Request. for Additional Information - Set 13", April 14, 2011.

In Reference 1, NextEra Energy Seabrook 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. provides NextEra Energy Seabrook response to the information requested in Reference 8 and provides revised response to RAI Follow-up B.2.1.27-2 requested in Reference

9. provides the revised NextEra Energy Seabrook PWR Vessel Internals Program. provides the revised AMR items for the Reactor Vessel Internals (Revised LRA Table 3.1.2-3). provides the revised 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. Commitment #1 has been revised as a result of response to the RAI related to the aging management of PWR Vessel Internals. This letter also makes editorial corrections to Commitments # 12, # 13, and #26. Corrections to commitments #12, #13 and #26 now reflect the changes that were previously submitted in SBK-L-14037, dated March 5, 2014.

The changes are explained, and where appropriate to facilitate understanding, portions of the LRA are repeated with the change highlighted by strikethroughs for deleted text and bolded italics for inserted text. In some instances the entire text of a section has been replaced or added.

In these cases a note is included in the introduction indicating the replacement of the entire text of the section.

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. Michael H. Ossing Licensing Manager, at (603) 773-7512.

U.S. Nuclear Regulatory Commission SBK-L-14089/Page 3 I declare under penalty of perjury that the foregoing is true and correct.

Executed on June 2_q, 2014.

Sincerely, NextEra Energy Seabrook, LLC Dean Curtland Site Vice President

Enclosures:

- NextEra Energy Seabrook Response to Request for Additional Information Related to the Review of the Seabrook Station License Renewal Application- Set 21 - Revised NextEra Energy Seabrook PWR Vessel Internals Program - Revised AMR Items for the Reactor Vessel Internals (Revised LRA Table 3.1.2-3) - LRA Appendix A - Final Safety Report Supplement Table A.3, License Renewal Commitment List Updated to Reflect Changes to Date cc: W. M. Dean NRC Region I Administrator J. G. Lamb NRC Project Manager P. C. Cataldo NRC Senior Resident Inspector R. A. Plasse NRC Project Manager, License Renewal L. M. James NRC Project Manager, License Renewal Mr. Perry Plummer 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 Mr. John Giarrusso, Jr., Nuclear Preparedness Manager The Commonwealth of Massachusetts Emergency Management Agency 400 Worcester Road Framingham, MA 01702-5399

Enclosure 1 to SBK-L-14089 NextEra Energy Seabrook Response to Request for Additional Information Related to the Review of the Seabrook Station License Renewal Application- Set 21 And Revised Response to RAI Follow-up B.2.1.27-2 Provided in SBK-L-11063 dated April 14, 2011

U.S. Nuclear Regulatory Commission SBK-L-14089/Enclosure 1/Page 2 RAI LRA Appendix B Aging Management of Reactor Vessel Internals

Background:

The license renewal application (LRA) for Seabrook Station proposed aging management for the reactor vessel internal (RVI) components based on a regulatory commitment in the LRA's Updated Final Safety Analysis Report (UFSAR) Supplement. The commitment stated that the applicant will develop an aging management program (AMP) and inspection plan based on augmented inspection activities for the components developed by the EPRI Materials Reliability Project (MRP), and that the inspection plan will be submitted for NRC review and approval not later than 2 years after receipt of the renewed license or not less than 24 months prior to the period of extended operation, whichever comes first.

The NRC's recommended AMP for Pressurized Water Reactor (PWR) RVIs in Revision 2 of NUREG- 1801, Generic Aging Lessons Learned (GALL) Report, is given in Chapter XI.M 16A, "PWR Vessel Internals," which was issued in December 2010. On January 9, 2012, subsequent to the issuance of Revision 2 of the GALL Report, the EPRI MRP issued Technical Report No.

1022863, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A)," which included the NRC safety evaluation (SE) for the report's methodology dated December 16, 2011. On June 3, 2013, the staff revised AMP XI.M16A and the aging management review (AMR) items in GALL for PWR RVI components to be consistent with the contents of the MRP-227-A report and issued them in License Renewal Interim Staff Guidance Document No. LR-ISG-2011-04, "Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors."

On July 21, 2011, the NRC issued Regulatory Information Summary (RIS) 2011-07, "License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management,"

which provided updated NRC procedures for LRA reviews of PWR RVI AMPs. This RIS identified Category C plants as those plants that have an LRA currently under review, and stated that these applicants will be expected to revise their commitment for aging management of PWR vessel internals such that the information identified in the SE for MRP-227 would be submitted to the NRC for review and approval not later than two years after issuance of the renewed license or not later than two years before the plant enters the PEO, whichever comes first. Seabrook Station is a Category C plant in accordance with the RIS.

Issue:

The categorization of Seabrook Station and other plants in Category C of RIS 2011-07 was based on an expectation that the LRA would be reviewed and approved on a normal review schedule of 22 months, and that it would be an unreasonable burden to expect the applicant to address all aspects of the NRC's SE on MRP-227 within the LRA review. Since the completion of the staffs review of the LRA is still ongoing due to the ASR open item, the staff has concluded that

U.S. Nuclear Regulatory Commission SBK-L-14089/Enclosure 1/Page 3 the applicant should provide an LRA update or amendment that includes updated AMP and AMR items for the RVI components, including responses to the applicable Applicant/License Actions Items identified in the staff's SE for MRP-227 dated December 16, 2011.

Request: The staff requests that the applicant provide either an LRA amendment or an update that includes updated AMP and AMR items for the PWR RVI components at the Seabrook Station that are based on the guidance in LR-ISG-2011-04, including responses to applicable Applicant/License Actions Items identified in the staff's SE for MRP-227 dated December 16, 2011.

NextEra Energy Seabrook Response to RAI LRA Appendix B Aging Management of Reactor Vessel Internals NextEra Energy Seabrook previously submitted a PWR Vessel Internals Aging Management Program (RVI AMP) as part of its License Renewal Application (Ref. 1). The RVI AMP was based on EPRI MRP-227 Rev. 0 (Pressurized Water Reactor Internals Inspection and Evaluation Guidelines), which was issued in December 2008 (Ref. 2). As part of the NextEra Energy Seabrook's License Renewal Application (LRA), a commitment was made to implement the program prior to the period of extended operation and to submit an inspection plan to the NRC not less than 24 months prior to the period of extended operation.

In SBK-L-1 1069 dated April 22, 2011, NextEra Energy Seabrook revised its commitment to submit the inspection plan to the NRC not later than 2 years after receipt of the renewed license or not less than 24 months prior to the period of extended operation, whichever came first. This revised commitment was consistent with the Regulatory Issue Summary (RIS 2011-07),

"License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management", which was issued by the NRC on July 21, 2011 (Ref. 3). NextEra Energy Seabrook fell into Category.C (plants that have a license renewal application currently under review) in accordance with the guidance provided in RIS-2011-07.

On January 12, 2009, MRP-227 Rev.0 was submitted to the NRC for review and approval through the Nuclear Energy Institute. On June 22, 2011, the NRC issued Revision 0 of the Safety Evaluation Report (SER) for MRP-227 (Ref. 4). On December 16, 2011, the NRC issued Revision 1 of the Safety Evaluation Report for the final version of MRP-227 [(i.e. MRP-227-A)(Ref. 6)].

In the SER, the NRC staff determined that MRP-227-A is acceptable for use in PWR Vessel Internals aging management program in License Renewal Applications. The SER includes eight plant-specific Applicant/Licensee Action Items. Applicant/Licensee Action Items 4 and 6 are only applicable to B&W plants. NextEra Energy Seabrook is a Westinghouse NSSS plant.

U.S. Nuclear Regulatory Commission SBK-L-14089/Enclosure l/Page 4 Therefore, Applicant/Licensee Action Items 4 and 6 are not applicable to NextEra Energy Seabrook.

On May 28, 2013, the NRC issued License Renewal Interim Staff Guidance LR-ISG-2011-04, "Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors" (Ref. 5). This LR-ISG revised the guidance provided in NUREG-1801, Revision 2, "Generic Aging Lessons Learned Report" for the aging management of PWR Reactor Vessel Internal components exposed to reactor coolant environment.

In response to RAI B.2.1.7, the following changes are made to NextEra Energy Seabrook's LRA:

1. PWR Vessel Internals Program described in LRA Section B.2.1.7 is revised as described in Enclosure 2.

Note: The entire text of the PWR Vessel Internals Program has been replaced.

2. The PWR Vessel Internals Inspection Plan for NextEra Energy Seabrook is provided in Tables 1 through 4 of the PWR Vessel Internals Aging Management Program.
a. Table 1 describes the RVI components that are classified as Existing Program components.
b. Table 2 describes the RVI components classified as Primary components, which is based on MRP-227-A, Table 4.3.
c. Table 3 describes the RVI components classified as Expansion components, which is based on MRP-227-A, Table 4.6.
d. Table 4 describes the Examination Acceptance and Expansion criteria for Westinghouse plants, which is based on MRP-227-A, Table 5.3.
3. AMR items for Reactor Vessel Internals (LRA Table 3.1.2-3) have been revised as described in Enclosure 3.
  • Note: The entire LRA Table 3.1.2-3 has been replaced.
4. In LRA Appendix A, Section A.2.1.7, PWR Vessel Internals, has been revised as follows.

Note: The entire text of LRA Section A.2.1.7 has been replaced.

The PWR Vessel Internals Program implements the guidance provided in EPRI Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A, Technical Report 1022863) and EPRI Inspection Standard for PWR Internals (MRP-228, Technical Report 10166609).

The program is a condition monitoring program designed to manage the aging effects on the PWR vessel internal components. The recommended activities provided in MRP-227-A and additional plant-specific activities not defined in MRP-227-A are implemented in accordance with Nuclear Energy Institute (NEI) 03-08, "Guideline for the Management of Materials Issues."

U.S. Nuclear Regulatory Commission SBK-L-14089/Enclosure 1/Page 5 This program is used to manage the effects of age-related degradation mechanisms that are applicable to the PWR vessel internal components. These aging effects include: a) cracking, including stress corrosion cracking (SCC), primary water stress corrosion cracking (PWSCC), irradiation-assisted stress corrosion cracking (IASCC), and cracking due to fatigue/cyclic loading, b) loss of material induced by wear, c) loss of fracture toughness due to either thermal aging or neutron irradiation embrittlement, d) changes in dimensions due to void swelling or distortion, and e) loss of preload due to thermal and irradiation-enhanced stress relaxation or creep.

5. Commitment # 1 is revised as follows to provide confirmation and acceptability of the implementation of MRP-227-A by addressing the plant-specific Applicant/Licensee Action Items outlined in section 4.2 of the NRC SER. NextEra Energy Seabrook is currently working with the NSSS supplier to establish a submittal schedule to address the plant-specific Applicant/Licensee Action Items outlined in section 4.2 of the NRC SER.

PROGRAM or UFSAR No. COMMITMENT SCHEDULE TOPIC LOCATION Program to be implemented A n____________p lan _for- _____ p rior-to th e p erio d o f ex ten d ed -

Vnsiln

.. peTi*Io pl fe R.b.ac*ti operation. Inspection plan to be foe- el Intevn li ll bedsubmite d subm itted to NRC not later-t a foradRC aproal.2 eviw yeas after-receipt of the PWR Vessel Provide confirmationand renewed license or not less than Internals acceptability of the implementation A.2..1.7 24 months pr- to the period of.

of MRP-22 7-A by addressingthe extended operation, whichever.

plant-specificApplicant/Licensee .e...es iis..

Action Items outlinedin section 4.2 NextEra Energy Seabrook will of the NRC SER. provide a submittalschedule by October 15, 2014 RAI B.2.1.22-6

Background:

The Buried Piping Inspection Locations table provided in a letter dated July 2, 2013, states the number of inspections based on the material type, interval of inspection, and status of cathodic protection, backfill, and soil sample results.

U.S. Nuclear Regulatory Cormnission SBK-L-14089/Enclosure 1/Page 6 Issue:

The staff has concluded that the referenced table is consistent with LR-ISG-2011-03 Table 4a, "Inspections of Buried Pipe," with the following exceptions:

1. The not to exceed number of inspections for polymeric piping in the 40-50 year period is three in the referenced table and four in Table 4a of LR-ISG-2011-03. The fire protection system is the only in-scope system with polymeric piping. The Buried Piping and Tanks Inspection Program states that it may conduct jockey pump monitoring in lieu of excavated direct visual inspections of fire protection piping. However, if the program were changed to use excavated direct visual inspections of the fire protection system in lieu of jockey pump monitoring, the number of inspections would not be consistent with LR-ISG-2011-03.
2. Footnote 5 of the referenced table states, "[i]f cathodic protection does not meet Category C and backfill has been determined to be inadequate, buried steel piping will be inspected as Category F [the highest inspection quantity category]." Footnote 2 of the referenced table states, "[t]he adequacy of backfill will be determined by the condition of coatings and base materials noted during inspections. If damage to the coatings or base materials is determined to have been caused by the backfill, the backfill will be considered to be 'inadequate' (for the purpose of this program)." The applicant's table states that Category F inspections will be performed if the soil is determined to be corrosive.

LR-ISG-2011-03 Table 4a states that the conditions that result in conducting the highest number of inspections (Category F) are: (a) coatings and backfill have not been provided in accordance with the "preventive actions" program element, (b) a leak has occurred in buried piping due to external corrosion, (c) significant coating degradation or metal loss in more than 10 percent of inspections conducted, and (d) the soil has been demonstrated to be corrosive for steel piping.

The staff concludes the following in relation to each of the Table 4a conditions:

a. The referenced table is not consistent with condition (b) because it has not stated leaks as a criterion for entry into Category F.
b. The referenced table is not consistent with condition (c) because, although footnote 2 states that if damage to the coatings or base materials are determined to have been caused by the backfill, the backfill will be considered to be inadequate (resulting in use of Category F to determine the number of inspections), the Table 4a condition is not solely based on damage to coatings by backfill.

Request:

1. State the basis why conducting three excavated direct visual inspections of polymeric piping in the 40-50 year period; in lieu of the jockey pump monitoring; is adequate to provide reasonable assurance that the piping will meet its current licensing basis intended function(s).

U.S. Nuclear Regulatory Commission SBK-L-14089/Enclosure l/Page 7 Alternatively, revise the table to be consistent with the number of recommended inspections for buried polymeric piping in LR-ISG-2011-03.

2. State the basis for why the proposed criteria for invoking Category F inspections will be adequate to provide reasonable assurance that the piping will meet its current licensing basis intended function(s). Alternatively, revise the table to be consistent with LR-ISG-201 1-03.

NextEra Energy Seabrook Response to RAI B.2.1.22-6 In response to RAI B.2.1.22-6, the "Buried Piping Inspection Locations Table" provided in SBK-L-13115 (dated July 2, 2013) and SBK-L-13183 (dated October 21, 2013) has been revised as follows:

Buried Piping Inspection Locations Inspections each Systems Currently Material Status of Cathodic GALL Curry 10-Year Period 1Systes in this Category Type Protection Category 30-40 40-50 50-60 AL6XN N/A N/A 0 0 0 None Stainless N/A N/A 1 1 1 CO, DG Steel 2 A Adequate Backfill 1 1 1 Inadequate Backfill 2"3 FP' Polymeric N/A B 1% 2% 3%

NTE 2 NTE-3 4 NTE 6 Installed, available and C 1 1I I effective 4 CBA, IA, FPSW External corrosion 1% 1% 1%

control not required NTE 2 NTE 2 NTE 2 Not practical, not Steel 5 installed, or installed E 5% 6% 7.5%

but not meeting Cat C; NTE 7 NTE 10 NTE 12 non-corrosive soil 6 AB 7, CBA, CO, DF, Not installed or DG, FW, FP8 installed but not 10% 12% 15%

meeting Cat C;6 NTE 15 NTE 20 NTE 25 corrosive soil GENERAL NOTES:

1. Each inspection will examine a minimum of 10 feet of pipe or the entire length of a run, whichever is less.
2. The adequacy of backfill will be determined by the condition of coatings and base materials noted during inspections. If damage to the coatings or base materials are determined to have been caused by the backfill, the backfill will be considered to be "inadequate" (for the purpose of this program).
3. If all polymeric pipe in-scope is non-safety related, the inspection quantities may be reduced by half.
4. Cathodic protection is available and effective if it

U.S. Nuclear Regulatory Commission SBK-L-14089/Enclosure I/Page 8

" was installed or refurbished 5 years prior to the end of the inspection period of interest; and

" has been operational (available) at least 85 percent of the time since 10 years prior to the PEO or since installation or refurbishment (exclusive of time off-line for testing), whichever is shorter; and

" has met the acceptance criteria of Section 6 at least 80 percent of the time since 10 years prior to the PEO or since installation or refurbishment, whichever is shorter.

5. if eathcdic pr-etcctien does not meet Category C, and baekfill has been determined to be inadequate, buried steel piping will ben.sp.eted as Catego.. F. Buriedpiping will be inspected as category F if; a) cathodic protection does not meet Category C, or b) significant coating degradation or metal loss in more than 10 percent of inspections conducted, or c) a leak has occurred in buried piping due to external corrosion,or d) the soil has been demonstratedto be corrosivefor steel piping.
6. Soil corrosivity is detennined by soil analysis using a demonstrated methodology such as EPRI report 1021470, Table 8-1. A soil corrosivity value of 10 or greater using this method is considered corrosive.
7. This line is not is use. It has been drained and flushed and is awaiting replacement per a design change. The inspection criteria for the replacement piping will be determined based on material selection, coating, cathodic protection, and quality of backfill.
8. If Fire Protection piping is inspected by excavation in lieu of by alternative testing (e.g., flow test, jockey pump monitoring), and the extent of examinations is not based on the percentage of piping in the material group, the Not-to-Exceed (NTE) value will be increased by 1 inspection, if normally less than 10, or 2 inspections, if normally 10 or greater.

RAI A.1-2, License Renewal Commitments and the UFSAR

Background:

By letter dated May 25, 2010, NextEra Energy Seabrook, LLC submitted an application pursuant to Title 10 of the Code of Federal Regulations (CFR) Part 54 to renew the operating license, NPF-86 for Seabrook Station, for review by the U.S. Nuclear Regulatory Commission (NRC) staff. The NRC staff is reviewing this application in accordance with the guidance in NUREG-1800, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants." During the review of the Seabrook license renewal application (LRA) by the NRC staff, NextEra Energy Seabrook made commitments related to aging management programs (AMPs),

aging management reviews (AMRs), and time-limited aging analyses, as applicable, related to managing the aging effects of structures and components prior to the PEO. The list of these commitments, as well as the implementation schedules and the sources for each commitment, will be included as a Table in Appendix A to the LRA and the SER.

In Section 1.7, "Summary of Proposed License Conditions," of the SER with Open Items, the staff stated that following its review of the LRA, including subsequent information and clarifications provided by the applicant, it identified proposed license conditions. The first license condition requires the information in the UFSAR supplement, submitted pursuant to

U.S. Nuclear Regulatory Commission SBK-L- 14089/Enclosure 1/Page 9 10 CFR 54.21 (d), as revised during the LRA review process, be made a part of the UFSAR.

The second license condition in part states that the new programs and enhancements to existing programs listed in Appendix A of the SER and the applicant's UFSAR supplement be implemented no later than 6 months prior to the PEO. This license condition also states, in part, that activities in certain other commitments shall be completed by 6 months prior to the PEO or the end of the last refueling outage prior to the PEO, whichever occurs later.

The NRC plans to revise Appendix A of the SER to align with this guidance, and to reformat the license condition to be as follows:

The UFSAR supplement submitted pursuant to 10 CFR 54.21 (d), as revised during the license renewal application review process, and as supplemented by Appendix A of NUREG [XXXX], "Safety Evaluation Report Related to the License Renewal of Seabrook Station" dated [Month Year], describes certain programs to be implemented and activities to be completed prior to the PEO.

a. The licensee shall implement those new programs and enhancements to existing programs no later than 6 months prior to the PEO.
b. The licensee shall complete those inspection and testing activities, as noted in Commitment Nos. [x] through [xx] of Appendix A of NUREG

[XXXX], by the 6 month date prior to PEO or the end of the last refueling outage prior to the PEO, whichever occurs later.

The licensee shall notify the NRC in writing within 30 days after having accomplished item (a) above and include the status of those activities that have been or remain to be completed in item (b) above.

The staff also notes that in the course of its evaluating multiple commitments to be implemented in the future in order to arrive at a conclusion of reasonable assurance that requirements of 10 CFR 54.29(a) have been met, these license renewal commitments must be incorporated either into a license condition or into a mandated licensing basis document, such as the UFSAR. Those commitments that are incorporated into the UFSAR are typically done so by incorporating each one verbatim (or by a summary and a commitment reference number) into the respective UFSAR summaries in the applicant's LRA Appendix A.

Issue:

As reflected in the SER Appendix A, the implementation schedule for some commitments may conflict with the implementation schedule intended by the generic license condition. In addition, these licensing commitments need to be incorporated either into a license condition or into the

U.S. Nuclear Regulatory Commission SBK-L-14089/Enclosure l/Page 10 applicant's UFSAR summary in such a manner as discussed above.

Request:

1. Identify those commitments to implement new programs and enhancements to existing programs. Indicate the expected date for completing the implementation of each of these programs and enhancements.
2. Identify those commitments to complete inspection or testing activities prior to the PEO.

Indicate the expected dates for the completion of each of these inspection and testing activities.

3. For each commitment provided by the applicant in the SER Appendix A, identify where and how NextEra Energy Seabrook, LLC proposes that it be incorporated: into either a license condition or into the Seabrook UFSAR.

NextEra Energy Seabrook Response to RAI A.1-2, License Renewal Commitments and the UFSAR In response to RAI A. 1-2, Requests 1, 2 & 3, the folowing has been added to the end of LRA Appendix A, A. 1 Introduction.

Commitments for implementing new programs and enhancements to existing programs with schedule dates as prior to the period of extended operationshall be completed no later than 6 months priorto the periodof extended operation.

Commitments for completion of inspection and testing activities shall be completed no later than 6 months prior to the period of extended operation or the last refueling outage prior to the periodof extended operation, whichever is later.

The final version of the License Renewal Commitment List will be included in the NextEra Energy Seabrook UFSAR Supplement (LRA Appendix A) before incorporationinto the NextEra Energy Seabrook UFSAR (after NRC approvalof the LRA). After incorporation into the NextEra Energy Seabrook UFSAR, changes to information within the UFSAR Supplement will be made in accordancewith 10 CFR 50.59.

NextEra Energy Seabrook will notify the NRC in writing within 30 days after having accomplished items listed in the License Renewal Commitment List and include the status of those activities that have been or remain to be completed.

U.S. Nuclear Regulatory Commission SBK-L-14089/Enclosure 1/Page 11 Revised NextEra Energy Seabrook Response to RAI Follow-up B.2.1.27-2 Provided in SBK-L-11063 dated April 14, 2011 In SBK-L- 11063 dated April 14, 2011 (Reference 9), NextEra Energy Seabrook provided a response to RAI Follow-up B.2.1.27-2 describing the planned actions related to monitoring liner plate thickness around the fuel transfer tube and efforts to address leakage of borated water.

At the time the response to the request for additional information was submitted, the liner plate was identified as being Category E-C in accordance with IWE-2420(b) which states in part "when examination results require evaluation of flaws or degradation in accordance with IWE-3000, and the component is acceptable for continued service or when the examination results in performance of a repair/replacement activity, the areas containing such flaws or area of degradation, or areas subject to a repair/replacement activity, shall be reexamined during the next inspection period listed in the schedule of the inspection program of IWE-24 11 or IWE-2412, in accordance with Table IWE-2500-1, Examination Category E-C." Upon subsequent review, it was determined that this area did not meet the requirements for Category E-C augmented inspections and was reclassified. The original coating indications and minor surface corrosion will not reoccur since the leakage into the vault was successfully remediated during OR 1I in October 2006. The areas of concern on the containment liner plate that were originally identified were examined and accepted in October 2009 (via IWE-VT-3 Examination and UT thickness measurements).

Based on the revised classification, NextEra Energy Seabrook has revised the original response to RAI B.2.1.27-2 as follows:

Request for Additional Information (RAI) Follow-up B.2.1.27-2

Background:

By letter dated December 17, 2010, the applicant responded to RAI B.2.1.27-2 and stated that the liner plate around the fuel transfer tube has been identified in the 1SI program for augmented inspection in accordance with the 1995 Edition with 1996 Addenda of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, Subsection IWE-2420(b) and (c).

Issue:

The ASME 1995 Edition with 1996 Addenda,Section XI, Subsection IWE-2420(b) and (c) states that reexamination of degraded areas is no longer required if these areas remains essentially unchanged for three consecutive inspection periods. However, it is not clear from the applicant's response if the containment liner plate around the fuel transfer tube is still exposed to the borated water leakage. Exposure to borated water can promote corrosion of the liner plate and adversely affect the ability of the liner to perform its intended function.

U.S. Nuclear Regulatory Commission SBK-L- 14089/Enclosure 1/Page 12 Request:

Describe steps that are being taken to monitor the liner plate thickness around the transfer tube and/or efforts to address the leakage of borated water.

Revised: NextEra Energy Seabrook Response to RAI Follow-up B.2.1.27-2:

The leak path into the fuel transfer- tube vault has been r-epair-ed and the bor.ated water- leakage stopped. The areas of the containment liner plate that had showed signs of deficiency (loss of UT e.xamintaions for- the t exam cyeles. if no ifiher- degadatien (loss of material) is observed dur-ing these three cycles, the subject area will return to normal visuial AVE inspections.

These visual inspections would be able to identify any futHher- leakage of bor-ated watr In SBK-L-11063 dated April 14, 2011, NextEra Energy Seabrookpreviously provided a response to RAI Follow-up B.2.1.27-2 describingthe plannedactions related to monitoring liner plate thickness aroundthe fuel transfertube and efforts to address leakage of borated water.

At the time the response to the requestfor additionalinformation was submitted, the liner plate was identified as being Category E-C in accordancewith IWE-2420(b) which states in part "when examination results requireevaluation offlaws or degradation in accordancewith IWE-3000, and the component is acceptablefor continued service or when the examination results in performance of a repair/replacementactivity, the areascontainingsuch flaws or area of degradation,or areassubject to a repair/replacementactivity, shall be reexamined during the next inspectionperiod listed in the schedule of the inspection program of IWE-2411 or IWE-2412, in accordancewith Table IWE-2500-1, Examination Category E-C."

Upon subsequentreview, it was determined that this area did not meet the requirementsfor Category E-C augmented inspections and was reclassified.The originalcoating indications and minor surface corrosion will not reoccur since the leakage into the vault was successfully remediated during ORll in October 2006. The areas of concern on the containmentliner plate that were originallyidentified were examined and accepted in October 2009 (via IWE-VT-3 Examination and UT thickness measurements).

Enclosure 2 to SBK-L-14089 Revised NextEra Energy Seabrook PWR Vessel Internals Program

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 2/ Page 2 B.2.1.7 PWR VESSEL INTERNALS PROGRAM Program Description The PWR Vessel Internals Program implements the guidance provided in EPRI Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A, Technical Report 1022863) and EPRI Inspection Standard for PWR Internals (MRP-228, Technical Report 10166609).

The program is a condition monitoring program designed to manage the aging effects on the PWR vessel internal components. The recommended activities provided in MRP-227-A and additional plant-specific activities not defined in MRP-227-A are implemented in accordance with Nuclear Energy Institute (NEI) 03-08, "Guideline for the Management of Materials Issues."

This program is used to manage the effects of age-related degradation mechanisms that are applicable to the PWR vessel internal components. These aging effects include: a) cracking, including stress corrosion cracking (SCC), primary water stress corrosion cracking (PWSCC),

irradiation-assisted stress corrosion cracking (IASCC), and cracking due to fatigue/cyclic loading, b) loss of material induced by wear, c) loss of fracture toughness due to either thermal aging or neutron irradiation embrittlement, d) changes in dimensions due to void swelling or distortion, and e) loss of preload due to thermal and irradiation-enhanced stress relaxation or creep.

The program applies the guidance provided in MRP-227-A, for inspecting, evaluating, and, if applicable, dispositioning non-conforming PWR vessel internal components. These examinations provide reasonable assurance that the effects of age-related degradation mechanisms will be managed during the period of extended operation. The program includes expanding periodic examinations and other inspections, if the extent of the degradation identified exceeds the expected levels.

MRP-227-A guidance for selecting RVI components for inclusion in the inspection sample is based on a four-step ranking process. Through this process, the PWR vessel internal components are categorized into four groups; 1) Primary, 2) Expansion, 3) Existing Programs, or 4) No Additional Measurements.

The result of this four-step sample selection process is a set of "Primary" PWR vessel internal component locations that are inspected because they are expected to show the leading indications of the degradation effects, with another set of "Expansion" internals component locations that are specified to expand the sample should the indications be more severe than anticipated.

The degradation effects in a third set of internals locations are deemed to be adequately managed by "Existing Programs," such as ASME Code,Section XI, B-N-3 examinations of core support structures. A fourth set of internals locations are deemed to require "No Additional Measures."

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 2/ Page 3 Program Elements The following provides the results of the evaluation of each program element against the 10 elements described in License Renewal Interim Staff Guidance LR-ISG-2011-04, "Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors".

Element I - Scope of Program The scope of the RVI AMP includes RVI components at NextEra Energy Seabrook, which is built to a Westinghouse design. The scope of the program applies the methodology and guidance in MRP-227-A, which provides augmented inspection and flaw evaluation methodology for assuring the functional integrity of safety-related internals in commercial operating U.S. PWR nuclear power plants designed by B&W, CE, and Westinghouse. The scope of components considered for inspection under MRP-227-A guidance includes core support structures (typically denoted as Examination Category B-N-3 by the ASME Code,Section XI), those RVI components that serve an intended license renewal safety function pursuant to criteria in 10 CFR 54.4(a)(1), and other RVI components whose failure could prevent satisfactory accomplishment of any of the functions identified in 10 CFR 54.4(a)(1)(i), (ii), or (iii).

The scope of the program does not include consumable items, such as fuel assemblies, reactivity control assemblies, and nuclear instrumentation, because these components are not typically within the scope of the components that are required to be subject to an aging management review. The scope of the program also does not include welded attachments to the internal surface of the reactor vessel because these components are considered to be ASME Code Class 1 appurtenances to the reactor vessel and are adequately managed in accordance with NextEra Energy Seabrook's ASME Code, Section Xl, Inservice Inspection, Subsections IWB, IWC, and IWD Program as described in Section B.2.1.1 of the LRA.

Element 2 - Preventive Actions The guidance in MRP-227-A relies on PWR water chemistry control to prevent or mitigate aging effects that can be induced by corrosive aging mechanisms such as loss of material induced by general corrosion, pitting corrosion, crevice corrosion, or stress corrosion cracking or any of its forms (SCC, PWSCC, or IASCC). NextEra Energy Seabrook's reactor coolant water chemistry is monitored and maintained in accordance with the Water Chemistry Program as described in Section B.2.1.2 of the LRA.

Additionally, NextEra Energy Seabrook replaced the original X-750 Guide Tube Assembly (GTA) support pins with cold worked (CW) 316 SS support pins during Refueling Outage 11 (Fall of 2006). This modification is considered a preventative action as the CW 316 SS support

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 2/ Page 4 pins are considered to be much more resistant to the degradation mechanisms of concern for the RVI components than the X-750 material. Therefore, there is no requirement for augmented inspections of the CW 316 SS GTA support pins under the PWR Vessel Internals Aging Management Program. However, appropriate actions will be taken upon receiving further recommendations from Westinghouse.

Element 3 - Parameters Monitored/Inspected A total of eight age related degradation mechanisms are considered applicable to the RVI components: 1) stress corrosion cracking (SCC), 2) irradiation assisted stress corrosion cracking (IASCC), 3) fatigue, 4) irradiation embrittlement (IE), 5) thermal embrittlement (TE), 6) wear,

7) void swelling (VS), and 8) irradiation enhanced stress relaxation/creep (ISR/IC). A brief description of these degradation mechanisms and the associated aging effects are as follows:

Stress CorrosionCracking (SCC)

SCC is a localized, non-ductile failure caused by a combination of stress, susceptible material, and an aggressive environment. The fracture path of SCC can be either transgranular or intergranular in nature. The aggressive contaminants most commonly associated with SCC of austenitic stainless steels are dissolved chlorides and oxygen. Nickel base alloys such as Alloy 600 and X-750 have exhibited susceptibility to intergranular SCC in primary water without the presence of aggressive contaminants, commonly referred to as primary water stress corrosion cracking (PWSCC). SCC of Stainless Steel (SS) in primary water is also considered feasible at high stress levels. The aging effect of SCC is cracking.

IrradiationAssisted SCC (IASCC)

IASCC is a form of intergranular SCC that results from the combined influence of neutron irradiation and an aggressive environment. A limited number of IASCC failures of RVI components, specifically fasteners, constructed of austenitic stainless steels and nickel base alloys have been observed. The aging effect of IASCC is cracking.

Fatigue Fatigue is defined as the structural deterioration that can occur as a result of the periodic application of stress by mechanical, thermal, or combined effects. High cycle fatigue results from relatively low cyclic stress (<yield strength) applied for many (>105) cycles. Low cycle fatigue results from relatively high cyclic stress (>yield strength) applied for low number of cycles. The aging effect of fatigue is cracking.

IrradiationEmbrittlement (IE)

IE refers to a gradual and progressive change in mechanical properties of a material resulting from exposure to high levels of neutron irradiation. These changes include an increase in yield and tensile strengths, and a corresponding decrease in ductility and toughness. The aging effect

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 2/ Page 5 of IE is loss of fracture toughness.

Thermal Embrittlement (TE)

Thermal embrittlement refers to the same gradual and progressive change in mechanical properties of a material as IE except it results from exposure to elevated temperatures rather than neutron irradiation. For the RVI components, TE is only a concern for SS castings and welds with duplex microstructures containing both ferrite and austenite. The aging effect of TE is loss of fracture toughness.

Wear Wear is caused by the relative motion between adjacent surfaces, with the extent determined by the relative properties of the adjacent materials and their surface condition. The aging effect of wear is loss of material.

Void Swelling (VS)

Void swelling is the gradual increase in volume of a component caused by the formation of microscopic cavities. These cavities result from the nucleation and growth of vacancies created by exposure to high levels of neutron irradiation. During the initial licensing periods of domestic PWRs, field experience has not revealed any evidence of VS in RVI components; however it is postulated as a possibility during periods of extended operation based upon accelerated laboratory testing. The aging effect of VS is dimensional change.

Irradiationand Thermally EnhancedStress Relaxation/Creep(SR/C)

Stress relaxation involves the short term unloading of preloaded components upon exposure to elevated temperatures or high levels of neutron irradiation. Creep is a longer term process in which plastic deformation occurs within a loaded component. The temperatures of RVI are typically not high enough to support creep; however it can develop upon exposure to high levels of neutron irradiation over an extended period. The aging effect of stress relaxation and creep is loss of preload.

For the management of cracking, the program monitors for evidence of surface breaking linear discontinuities if a visual inspection technique is used as the non-destructive examination (NDE) method, or for relevant flaw presentation signals if a volumetric UT method is used as the NDE method. For the management of loss of material, the program monitors for gross or abnormal surface conditions that may be indicative of loss of material occurring in the components. For the management of loss of preload, the program monitors for gross surface conditions that may be indicative of loosening in applicable bolted, fastened, keyed, or pinned connections. The program does not directly monitor for loss of fracture toughness that is induced by thermal aging or neutron irradiation embrittlement. Instead, the impact of loss of fracture toughness on component integrity is indirectly managed by using visual or volumetric examination techniques to monitor for cracking in the components and by applying applicable

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 2/ Page 6 reduced fracture toughness properties in the flaw evaluations if cracking is detected in the components and is extensive enough to warrant a supplemental flaw growth or flaw tolerance evaluation under the MRP-227-A guidance or ASME Code,Section XI requirements. The program uses physical measurements to monitor for any dimensional changes due to void swelling, irradiation growth, distortion, or deflection.

Specifically, the program implements the parameters monitored/inspected criteria and bases for inspecting the relevant parameter conditions for Westinghouse designed Primary Components in Table 4-3 of MRP-227-A and for Westinghouse designed Expansion Components in Table 4-6 of MRP-227-A.

Element 4 - Detection of Aging Effects The RVI components have been categorized as Existing Program, Primary, Expansion, or No Additional Measurements, based on the guidance provided in MRP-227-A. A description of the component categories is as follows:

Existing Program Components:

Existing Program Components are susceptible to at least one of the eight degradation mechanisms, for which existing plant programs are capable of managing the associated aging effect(s). Details of the required inspections for Existing Program Components are provided in Table 1, Westinghouse Plants Existing Program Components Applicable to NextEra Energy Seabrook.

PrimaryComponents:

Primary Components are highly susceptible to at least one of the eight degradation mechanisms, for which augmented inspections are required on a periodic basis to manage the associated aging effect(s). Primary Components are considered lead indicators for the onset of the applicable degradation mechanism(s). Details of the required inspections for Primary Components are provided in Table 2, Westinghouse Plants Primary Components Applicable to NextEra Energy Seabrook. The Primary group also includes components which have shown a degree. of tolerance to a specific aging degradation effect, but for which no highly susceptible component exists or for which no highly susceptible component is accessible.

Expansion Components:

Expansion Components are highly or moderately susceptible to at least one of the eight degradation mechanisms, but for which functionality assessment has shown a degree of tolerance to those aging effects. Augmented inspections are required once a specified level of degradation is detected in a linked Primary Component. Details of the required inspections for Expansion Components are provided in Table 3, Westinghouse Plants Expansion Components Applicable to

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 2/ Page 7 NextEra Energy Seabrook.

No Additional Measures Components.

No Additional Measures Components are either not susceptible to any of the eight degradation mechanisms, or if susceptible, the impact of failure on the functionality of the RVI components is insignificant. No further action is required for managing the aging of these RVI components.

Proven inspection methodologies are utilized to detect evidence of the relevant aging mechanism(s) for the Existing Programs, Primary, and Expansion Components. These include the following:

a) Direct physical measurements to monitor for loss of material or preload.

b) VT-3 (visual) exams to monitor for general degradation associated with loss of material or preload.

c) EVT- 1 (enhanced visual) exams to monitor for surface breaking linear discontinuities indicative of cracking.

d) UT (ultrasonic) exams to monitor directly for cracking.

e) ECT (eddy current testing) to further characterize conditions detected by VT-3 and EVT- 1 examination.

The requirements for the inspection methodologies and qualification of NDE systems used to perform those inspections are provided in EPRI MRP-228, "Inspection Standard for PWR Internals".

In some cases (as defined in MRP-227-A), physical measurements are used as supplemental techniques to manage for the gross effects of wear, loss of preload due to stress relaxation, or for changes in dimension due to void swelling, deflection or distortion. The physical measurement methods applied in accordance with this program include measurement of hold down spring height.

Inspection coverage for "Primary" and "Expansion" RVI components is implemented consistent with Sections 3.3.1 and 3.3.2 of the NRC SE, Revision 1, on MRP-227.

Element 5 - Monitoring and Trending The methods for monitoring, recording, evaluating, and trending the data that result from the program's inspections are given in Section 6 of MRP-227-A and its subsections. The evaluation methods include recommendations for flaw depth sizing and for crack growth determinations as well as for performing applicable limit load, linear elastic and elastic-plastic fracture analyses of relevant flaw indications. The examinations and re-examinations required by the MRP-227-A guidance, together with the requirements specified in MRP-228 for

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 2/ Page 8 inspection methodologies, inspection procedures, and inspection personnel, provide timely detection, reporting, and corrective actions with respect to the effects of the age-related degradation mechanisms within the scope of the program.

The extent of the examinations, beginning with the sample of susceptible P W R internals component locations identified as Primary Component locations, with the potential for inclusion of Expansion Component locations if the effects are greater than anticipated, plus the continuation of the Existing Programs activities, such as the ASME Code,Section XI, Examination Category B-N-3 examinations for core support structures, provides a high degree of confidence in the total program.

Inspection Frequencies The required inspection frequencies for Existing, Primary and Expansion Components are specified in Tables 1, 2 and 3, respectively. Specified inspection frequencies are considered adequate to manage aging effects. However more frequent inspections may be warranted based upon an internal and external OE.

Inspection Coverage The required inspection coverage for Primary and Expansion Components are specified in Tables 2 and 3, respectively. The required inspection coverage for the Existing Program Components is as specified in the applicable program document (e.g. ASME Section XI). If the specified coverage for any of these components cannot be obtained, the condition shall be addressed in the Corrective Action Program (CAP).

Element 6 - Acceptance Criteria The acceptance criteria for Primary and Expansion Components are provided in Table 4, Westinghouse Plants Examination Acceptance and Expansion Criteria Applicable to NextEra Energy Seabrook. These criteria are based upon the requirements of ASME Section XI. All detected relevant conditions must be addressed in the CAP prior to plant start-up. Possible disposition options include: 1) supplemental exams to further characterize a detected condition,

2) engineering evaluation for continued service until the next inspection, 3) repair, or 4) replacement.

Engineering evaluations for continued service shall be conducted in accordance with NRC approved methodologies. WCAP-1 7096, "Reactor Internals Acceptance Criteria Methodology and Data Requirements", is currently under NRC review for this purpose. The potential loss of fracture toughness must be considered in any flaw evaluations.

Expansion Criteria The expansion criteria for expanding the scope of examination from the Primary to the linked Expansion Components, including the timing of inspections, are provided in Table 4,

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 2/ Page 9 Westinghouse Plants Examination Acceptance and Expansion Criteria.

It should be noted that the categorizations and associated inspection requirements described above do not replace or relieve any of the current ASME Section XI inspection requirements for the RVI components.

Element 7 - Corrective Actions Corrective actions following the detection of unacceptable conditions are fundamentally provided for in the NexEra Energy Seabrook Corrective Action Program. Any detected conditions that do not satisfy the examination acceptance criteria are required to be dispositioned through the plant corrective action program, which may require repair, replacement, or analytical evaluation for continued service until the next inspection. The disposition will ensure that design basis functions of the reactor internals components will continue to be fulfilled for all licensing basis loads and events. Examples of methodologies that can be used to analytically disposition unacceptable conditions are found in the ASME Code,Section XI or in Section 6 of MRP-227-A. Section 6 of MRP-227-A describes the options that are available for disposition of detected conditions that exceed the examination acceptance criteria of Section 5 of the report.

These include engineering evaluation methods conducted in accordance with WCAP- 17096, as well as supplementary examinations to further characterize the detected condition, or the alternative of component repair and replacement procedures. The latter are subject to the requirements of the ASME Code,Section XI. The implementation of the guidance in MRP-227-A, plus the implementation of any ASME Code requirements, provides an acceptable level of aging management of safety-related components addressed in accordance with the corrective actions of 10 CFR Part 50, Appendix B, as applicable.

Element 8 - Confirmation Process NextEra Energy Seabrook quality assurance procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR Part 5 0, Appendix B, as applicable. It is expected that the implementation of the guidance in MRP-227-A will provide an acceptable level of quality for inspection, flaw evaluation, and other elements of aging management of the PWR internals that are addressed in accordance with the 10 CFR Part 50, Appendix B (as applicable), confirmation process, and administrative controls.

Element 9 - Administrative Controls The administrative controls for License Renewal RVI AMP including the implementing procedures, review and approval processes, are under existing station 10 CFR 50 Appendix B Quality Assurance Programs. The PWR Vessel Internals AMP is established with a sufficient level of documentation and administrative controls to ensure effective long-term implementation.

The implementing procedure for NextEra Energy Seabrook is Chapter 3 of SASR (Seabrook Station RCS Materials Degradation Management Reference Manual), "Reactor Vessel Internals

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 2/ Page 10 Aging Management Program".

Element 10 - Operating Experience Few incidents of PWR internals aging degradation have been reported in operating U.S.

commercial PWR plants. A summary of observations to date is provided in Appendix A of MRP-227-A. Operating experience gained through Industry groups such as the EPRI MRP, the PWROG, INPO, WANO and International Sites shall be evaluated and incorporated into this program as needed in a timeframe consistent with the significance. Operating experience (OE) reports are continuously reviewed by NextEra Energy Seabrook personnel to ensure relevant OE is reviewed for impact on aging effects and/or aging management programs.

NUREG-1801 Consistency The NextEra Energy Seabrook PWR Vessel Internal Program is consistent with NUREG-1801 XI.M16A as modified by LR-ISG-2011-04, "Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors".

Exceptions to NUREG-1801 None Enhancements None Conclusion The NextEra Energy Seabrook PWR Vessel Internals Program provides reasonable assurance that the aging effects will be adequately managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.

Tables Table 1 - Westinghouse Plants Existing Programs Components Applicable to Seabrook Table 2 - Westinghouse Plants Primary Components Applicable to Seabrook Table 3 - Westinghouse Plants Expansion Components Applicable to Seabrook Table 4 - Westinghouse Plants Examination, Acceptance, and Expansion Criteria

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 2/ Page 11 Table 1 - Westinghouse Plants Existing Program Components Applicable to NextEra Energy Seabrook (Sheet 1 of 1)

Component Applicability Effect Reference Document Generic Requirement Examination Method and (Note 2) (Mechanism) Description Frequency Core Barrel Assembly All plants Loss of material ASME Code Visual (VT-3) examination to All accessible surfaces; one time per Core barrel flange (wear) Section XI determine general condition for interval excessive wear Upper Internals Assembly All plants Cracking (SCC, ASME Code Visual (VT-3) examination All accessible surfaces; one time per Upper support ring or skirt fatigue) Section XI interval Lower Internals Assembly All plants Cracking (SCC, ASME Code Visual (VT-3) examination of All accessible surfaces; one time per Lower core plate IASCC, fatigue) Section XI the lower core plates to detect interval evidence of distortion and/or loss of bolt integrity Lower Internals Assembly All plants Loss of material ASME Code Visual (VT-3) examination All accessible surfaces; one time per Lower core plate (wear) Section XI interval Bottom Mounted Loss of material BMI-FTT-IP Surface (ET) examination ET surface examination of full length Instrumentation System Not (wear) tubes at frequency specified in Flux thimble tubes Applicable BMI-FTT-IP. Tube selection and frequency based upon engineering (Note 3) evaluation of previous examination results Alignment and Interfacing All plants Loss of material ASME Code Visual (VT-3) examination All accessible surfaces; one time per Components (wear) Section XI interval Clevis insert bolt (Note 1)

Alignment and Interfacing All plants Loss of material ASME Code Visual (VT-3) examination All accessible surfaces; one time per Components (wear) Section XI interval Upper core plate alignment pins Notes:

1. Bolt was screened in because of stress relaxation and associated cracking; however, wear of the clevis/insert is the issue. Note doesn't apply to lower core plates.
2. There is no formal Program document for the GTA Support Pin Replacement Program; however, appropriate actions will be taken upon receiving further recommendation from Westinghouse. This program is not listed in this table because it does not include inspections of any RVI components.
3. NextEra Energy Seabrook does not utilize a Flux Thimble Tube Inspection Program because of the unique double wall design of the flux thimble tubes.

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 2/ Page 12 Table 2 - Westinghouse Plants Primary Components Applicable to NextEra Energy Seabrook (Sheet I of 4)

Effect Expansion Link Examination. Examination Component Applicability (Mechanism) (Note 1) Method Coverage Control Rod Guide All plants Loss of material (wear) None Visual (VT-3) examination no 20% examination of the number of CRGT Tube Assembly later than 2 refueling outages assemblies, with all guide cards within Guide plates (cards) from the beginning of the license each selected CRGT assembly examined renewal period, and no earlier than two refueling outages prior to the start of the license renewal period. Subsequent examinations are required on a ten-year interval Control Rod Guide All plants Cracking (SCC, Bottom-mounted Enhanced visual (EVT-1) 100% of outer (accessible) lower flange Tube Assembly fatigue) instrumentation examination to determine the weld surfaces and adjacent base metal on Lower flange Welds Aging Management column bodies, presence of crack-like surface the individual peripheral assemblies (IE and TE) Lower support flaws in flange welds no later (Note 2) colunm bodies than 2 refueling outages from the (cast), Upper core beginning of the license renewal plate, Lower period and subsequent support examination on a ten-year casting/forging interval Core Barrel Assembly All plants Cracking (SCC) Lower support Periodic enhanced visual 100% of one side of the accessible Upper core barrel flange column bodies (EVT-1) examination, with surfaces of the selected weld and adjacent weld (non-cast) 10-year intervals, no later than base metal (Note 4)

Core barrel outlet 2 refueling outages from the nozzle welds beginning of the license renewal period and subsequent examination on a ten-year interval Core Barrel Assembly All plants Cracking (SCC, Upper and lower Periodic enhanced visual 100% of one side of the accessible Upper and lower core IASCC, Fatigue) core barrel cylinder (EVT-1) examination, with surfaces of the selected weld and adjacent barrel cylinder girth axial welds 10-year intervals, no later than base metal (Note 4) welds 2 refueling outages from the beginning of the license renewal period and subsequent examination on a ten-year interval

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 2/ Page 13 Table 2 - Westinghouse Plants Primary Components Applicable to NextEra Energy Seabrook (Sheet 2 of 4)

Effect Expansion Link Examination Examination Component Applicability (Mechanism) (Note 1) Method Coverage Core Barrel Assembly All plants Cracking (SCC, None Periodic enhanced visual 100% of one side of the accessible Lower core barrel flange Fatigue) (EVT-1) examination, with surfaces of the selected weld and adjacent weld (Note 5) 10-year intervals, no later than base metal (Note 4) 2 refueling outages from the beginning of the license renewal period and subsequent examination on a ten-year interval Baffle-Former All plants with Cracking (IASCC, None Visual (VT-3) examination, with Bolts and locking devices on high fluence Assembly baffle-edge fatigue) that results in baseline examination between seams. 100% of components accessible Baffle-edge bolts bolts

  • Lost or broken 20 and 40 EFPY and subsequent from core side (Note 3) gdevices examinations on a ten-year locking dinterval
  • Failed or missing bolts

Baffle-Former All plants Cracking (IASCC, Lower support Baseline volumetric (UT) 100% of accessible bolts (Note 3). Heads Assembly fatigue) column bolts, examination between 25 and accessible from the core side. UT Baffle-former bolts Aging Management Barrel-former bolts 35 EFPY, with subsequent accessibility may be affected by (IE and ISR) (Note 6) examination on a ten-year complexity of head and locking device interval designs

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 2/ Page 14 Table 2 - Westinghouse Plants Primary Components Applicable to NextEra Energy Seabrook (Sheet 3 of 4)

Effect Expansion Link Examination Examination Component Applicability (Mechanism) (Note 1) Method Coverage Baffle-Former All plants Distortion (void None Visual (VT-3) examination to Core side surface as indicated Assembly swelling), or cracking check for evidence of distortion, Assembly (Includes (IASCC) that results in with baseline examination baffle plates, baffle edge

  • Abnormal interaction between 20 and 40 EFPY and bolts and indirect effects with fuel assemblies subsequent examinations on a of void swelling in ten-year interval former plates)
  • Gaps along high fluence baffle joint 0 Vertical displacement of baffle plates near high fluence joint e Broken or damaged edge bolt locking systems along high fluence baffle joint Alignment and All plants with Distortion (loss of load) None Direct measurement of spring Measurements should be taken at several Interfacing Type 304 Note: This mechanism height within three cycles of the points around the circumference of the Components stainless steel was not strictly beginning of the license renewal spring, with a statistically adequate Internals hold down hold down identified in the original period. If the first set of number of measurements at each point to
  • spring springs list of age-related degradation measurements is not sufficient to eliminate uncertainty. Replacement of determine life, spring height Type 304 springs by Type 403 springs is mechanisms measurements must be taken required when the spring stiffness is during the next two outages, in determined to relax beyond design order to extrapolate the expected tolerance spring height to 60 years

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 2/Page 15 Table 2 - Westinghouse Plants Primary Components Applicable to NextEra Energy Seabrook (Sheet 4 of 4)

Effect Expansion Link Examination Examination Component Applicability (Mechanism) (Note 1) Method Coverage Thermal Shield All plants with Cracking (fatigue) None Visual (VT-3) examination no 100% of thermal shield flexures Flexures thermal shields Loss of material (wear) later than 2 refueling outages that results in thermal from the beginning of the license shield flexures renewal period. Subsequent excessive wear, examinations on a ten-year fracture, or complete interval separation Notes:

1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 4. (General note applies to entire table).
2. A minimum of 75% of the total identified sample population must be examined.
3. A minimum of 75% of the total population (examined + unexamined), including coverage consistent with the Expansion criteria in Table 4, must be examined for inspection credit.
4. A minimum of 75% of the total weld length (examined + unexamined), including coverage consistent with the Expansion criteria in Table 4, must be examined from either the inner or outer diameter for inspection credit.
5. The lower core barrel flange weld may be alternatively designated as the core barrel-to-support plate weld in some Westinghouse plant designs.
6. Void swelling effects on this component is managed through management of void swelling on the entire baffle-former assembly.

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 2/ Page 16 Table 3 - Westinghouse Plants Expansion Components Applicable to NextEra Energy Seabrook (Sheet I of 2)

Effect Component Applicability (Mechanism) Primary Link Examination Method Examination Coverage Upper Internals Assembly All plants Cracking (fatigue, CRGT lower Enhanced visual (EVT- 1) 100% of accessible surfaces Upper core plate wear) flange weld examination (Note 2)

Aging Re-inspection every 10 years Management (IE) following initial inspection Lower Internals Assembly All plants Cracking CRGT lower Enhanced visual (EVT-1) 100% of accessible surfaces Lower support forging or Aging flange weld examination (Note 2) casting Management (TE Re-inspection every 10 years in casting) following initial inspection Core Barrel Assembly All plants Cracking (IASCC, Baffle-former Volumetric (UT) examination. 100% of accessible bolts.

Barrel-former bolts fatigue) bolts Re-inspection every 10 years Accessibility is limited by presence Aging following initial inspection of thermal shields or neutron pads Management (IE, (Note 2) void swelling, and ISR) ,

Lower Support Assembly All plants Cracking (IASCC, Baffle-former Volumetric (UT) examination. 100% of accessible bolts or as Lower support column bolts fatigue) bolts Re-inspection every 10 years supported by plant specific Aging following initial inspection justification (Note 2)

Management (IE and ISR)

Core Barrel Assembly All plants Cracking (SCC, Upper core Enhanced visual (EVT-1) 100% of one side of the accessible Core barrel outlet nozzle fatigue) barrel flange examination surfaces of the selected weld and welds Aging weld Re-inspection every 10 years adjacent base metal (Note 2)

Management (IE of following initial inspection lower sections)

Core Barrel Assembly All plants Cracking (SCC, Upper and Enhanced visual (EVT-1) 100% of one side of the accessible Upper and lower core barrel fatigue) lower core examination surfaces of the selected weld and cylinder axial welds Aging barrel cylinder Re-inspection every 10 years adjacent base metal (Note 2)

Management (IE) girth welds following initial inspection

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 2/ Page 17 Table 3 - Westinghouse Plants Expansion Components Applicable to NextEra Energy Seabrook (Sheet 2 of 2)

Component Applicability Effect Primary Link Examination Method Examination Coverage (Mechanism)

Lower Support Assembly All plants Cracking (IASCC) Upper core Enhanced visual (EVT-1) 100% of accessible surfaces Lower support column bodies Aging barrel flange examination (Note 2)

(non-cast) Management (IE) weld Re-inspection every 10 years following initial inspection Lower Support Assembly All plants Cracking (IASCC) Control rod Visual (EVT- 1) examination. 100% of accessible support Lower support column bodies including the guide tube Re-inspection every 10 years columns (Note 2)

(cast) detection of (CRGT) lower following initial inspection fractured support flanges columns Aging Management (IE)

Bottom Mounted All plants Cracking (fatigue) Control rod Visual (VT-3) examination of BMI 100% of BMI column bodies for Instrumentation System including detection guide tube column bodies as indicated by which difficulty is detected during of completely (CRGT) lower difficulty of insertion/withdrawal flux thimble insertion/withdrawal Bottom-mounted instrumentation (BMI) column fractured column flanges of flux thimbles bodies bodies. Re-inspection every 10 years Aging following initial inspection Management (IE) Flux thimble insertion/withdrawal to be monitored at each inspection interval Notes:

1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 4.
2. A minimum of 75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions).

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 2/Page 18 Table 4 - Westinghouse Plants Examination, Acceptance, and Expansion Criteria Applicable to NextEra Energy Seabrook (Sheet 1 of 6)

Item Applicability Examination Acceptance Additional Examination Criteria (Note 1) Expansion Link(s) Expansion Criteria Acceptance Criteria Control Rod Guide Tube All plants Visual (VT-3) examination None N/A N/A Assembly Guide plates (cards) The specific relevant condition is wear that could lead to loss of control rod alignment and impede control assembly insertion Control Rod Guide Tube All plants Enhanced visual (EVT-1) a. Bottom-mounted a. Confirmation of a. For BMI column bodies, Assembly examination instrumentation surface-breaking indications in the specific relevant Lower flange welds (BMI) column two or more CRGT lower condition for the VT-3 bodies flange welds, combined with examination is The specific relevant flux thimble completely fractured condition is a detectable insertion/withdrawal difficulty, column bodies crack-like surface b. Lower support shall require visual (VT-3) indication colunm bodies examination of BMI column (cast), upper core bodies by the completion of the b. For cast lower support plate and lower next refueling outage column bodies, upper support forging or core plate and lower casting support forging/casting,

b. Confirmation of the specific relevant surface-breaking indications in condition is a detectable two or more CRGT lower crack-like surface flange welds shall require indication EVT-1 examination of cast lower support column bodies, upper core plate and lower support forging/casting within three fuel cycles following the initial observation

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 2/ Page 19 Table 4 - Westinghouse Plants Examination, Acceptance, and Expansion Criteria Applicable to NextEra Energy Seabrook (Sheet 2 of 6)

Item Applicability Examination Acceptance Ex Crter i Additional Examination Criteria (Note 1) Expansion inks) xpansion rieria Acceptance Criteria Core Barrel Assembly All plants Periodic enhanced visual a. Core barrel outlet a. The confirmed detection and a and b. The specific Upper core barrel flange (EVT- 1) examination nozzle welds sizing of a surface-breaking relevant condition for the weld indication with a length greater core barrel outlet nozzle than two inches in the upper weld and lower support The specific relevant b. Lower support core barrel flange weld shall column body examination condition is a detectable column bodies (non- require that the EVT- I is a detectable crack-like crack-like surface cast) examination, and any surface indication indication supplementary UT examination, be expanded to include the core barrel outlet nozzle welds by the completion of the next refueling outage

b. If extensive cracking in the core barrel outlet nozzle welds is detected, EVT-1 examination shall be expanded to include the upper six inches of the accessible surfaces of the non-cast lower support column bodies within three fuel cycles following the initial observation

United States Nuclear Regulatory Commission SBK-L- 14089 / Enclosure 2/ Page 20 Table 4 - Westinghouse Plants Examination, Acceptance, and Expansion Criteria Applicable to NextEra Energy Seabrook (Sheet 3 of 6)

Examination Acceptance Additional Examination Item Applicability Criteria (Note 1) Expansion Link(s) Expansion Criteria Acceptance Criteria Core Barrel Assembly All Plants Periodic enhanced visual None None None Lower core barrel flange (EVT-I) examination weld (Note 2)

The specific relevant condition is a detectable crack-like surface indication Core Barrel Assembly All Plants Periodic enhanced visual Upper core barrel The confirmed detection and The specific relevant Upper core barrel cylinder (EVT-1) examination cylinder axial welds sizing of a surface breaking condition for the expansion girth welds indication with a length greater upper core barrel cylinder than two inches in the upper core axial weld examination is a The specific relevant barrel cylinder girth welds shall detectable crack-like condition is a detectable require that the EVT-I surface indication crack-like surface examination be expanded to indication include the upper core barrel cylinder axial welds by the completion of the next refueling outage Core Barrel Assembly All Plants Periodic enhanced visual Lower core barrel The confirmed detection and The specific relevant Lower core barrel cylinder (EVT-1) examination cylinder axial weld sizing of a surface breaking condition for the expansion girth welds indication with a length greater lower core barrel cylinder than two inches in the lower core axial weld examination is a The specific relevant barrel cylinder girth welds shall detectable crack-like condition is a detectable require that the EVT-I surface indication crack-like surface examination be expanded to indication include the lower core barrel cylinder axial welds by the completion of the next refueling outage

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 2/Page 21 Table 4 - Westinghouse Plants Examination, Acceptance, and Expansion Criteria Applicable to NextEra Energy Seabrook (Sheet 4 of 6)

Examination Acceptance Additional Examination Item Applicability Criteria (Note 1) Expansion Link(s) Expansion Criteria Acceptance Criteria Baffle-Former Assembly All plants with Visual (VT-3) examination None N/A N/A Baffle-edge bolts baffle-edge bolts The specific relevant conditions are missing or broken locking devices, failed or missing bolts, and protrusion of bolt heads Baffle-Former Assembly All plants Volumetric (UT) a. Lower support a. Confirmation that more than a and b. The examination Baffle-former bolts examination column bolts 5% of the baffle-former bolts acceptance criteria for the actually examined on the four UT of the lower support baffle plates at the largest colunm bolts and the The examination b. Barrel-forner bolts distance from the core barrel-former bolts shall be acceptance criteria for the (presumed to be the lowest dose established as part of the UT of the baffle-former locations) contain unacceptable examination technical bolts shall be established as indications shall require UT justification part of the examination examination of the lower technical justification support column bolts within the next three fuel cycles

b. Confirmation that more than 5% of the lower support column bolts actually examined contain unacceptable indications shall require UT examination of the barrel-former bolts

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 2/ Page 22 Table 4 - Westinghouse Plants Examination Acceptance and Expansion Criteria Applicable to NextEra Energy Seabrook (Sheet 5 of 6)

Item Applicability Examination Acceptance Additional Examination Criteria (Note 1) Expansion Link(s) Expansion Criteria Acceptance Criteria Baffle-Former Assembly All plants Visual (VT-3) examination None N/A N/A Assembly The specific relevant conditions are evidence of abnormal interaction with fuel assemblies, gaps along high fluence shroud plate joints, vertical displacement of shroud plates near high fluence joints, and broken or damaged edge bolt locking systems along high fluence baffle plate joints Alignment and Interfacing All plants with Direct physical None N/A N/A Components 304 stainless measurement of spring Internals hold down spring steel hold height down springs The examination acceptance criterion for this measurement is that the remaining compressible height of the spring shall provide hold-down forces within the plant-specific design tolerance

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 2/ Page 23 Table 4 - Westinghouse Plants Examination Acceptance and Expansion Criteria Applicable to NextEra

.Energy Seabrook (Sheet 6 of 6)

Examination Acceptance Additional Examination Item Applicability Criteria (Note 1) Expansion Link(s) Expansion Criteria Acceptance Criteria Thermal Shield Assembly All plants with Visual (VT-3) examination None N/A N/A Thermal shield flexures thermal shields The specific relevant conditions for thermal shield flexures are excessive wear, fracture, or complete separation Notes:

1. The examination acceptance criterion for visual examination is the absence of the specified relevant condition(s).
2. The lower core barrel flange weld may alternatively be designated as the core barrel-to-support plate weld in some Westinghouse plant designs.

Enclosure 3 to SBK-L-14089 Revised AMR Items for the Reactor Vessel Internals (Revised LRA Table 3.1.2-3)

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 3/ Page 2 Table 3.1.2-3 REACTOR VESSEL INTERNALS Summary of Aging Management Evaluation Aging Effect Requiring Aging NUREG Table Intended Componunctiond Material Environment Mag nagEfetReqint Management 1801 Vol. 3.X.1 Note Component Type Function Management Program 2 Item Item Alignment and Interfacing Structural Stainless Reactor Coolant. PWR Vessel IV.B2-33 components: internals hold down and Neutron Loss of Preload 3.1.1-27 A, 1 spring Support Steel Flux Intemals (R-108)

Alignment and Interfacing Structural Stainless Reactor Coolant PWR Vessel components: internals hold down and Neutron Changes in Dimensions IntePrals None None A, 1 spring Support Steel Flux Alignment and Interfacing Structural Stainless Reactor Coolant PWR Vessel components: interals hold down and Neutron Loss of Material V None None A, spring Support Steel Flux Internals PWR Vessel Alignment and Interfacing Structural Stainless Reactor Coolant Internals IV.B2-40 components: upper core plate Support Steel and Neutron Cracking (R-112) 3.1.1-37 alignment pins Flux Water Chemistry A, I Alignment and Interfacing Structural Stainless Reactor Coolant PWR Vessel IV.B2-34 components: upper core plate Support Steel and Neutron Loss of Material Internals (R- 115) 3.1.1-63 A, 1 alignment pins Flux Reactor Coolant PWR Vessel A, 1 Baffle-to- Former Assembly: Structural Stainless and Neutron Cracking Internals IV.B2-10 3.1.1-30 baffle-to-former bolts Support Steel Flux Water Chemistry (R-125)A, Baffle-to- Former Assembly: Structural Stainless Reactor Coolant PWR Vessel IV.B2-6 Baffle-to-Former tss l Structu Stanle and Neutron Loss of Fracture Toughness Internals (R-128) 3.1.1-22 A, I baffle-to-former bolts Support Steel Flux Baffle-to- Former Assembly: Structural Stainless Reactor Coolant PWR Vessel IV.B2-4 Support Steel and Neutron Change in Dimensions Internals (R- 126) 3.1.1-33 A, 1 baffle-to-former bolts Flux Reactor Coolant PWR Vessel IV.B2-5 Baffle-to- Former Assembly: Structural Stainless and Neutron Loss of Preload Internals (R- 129) 3.1.1-27 A, 1 baffle-to-former bolts Support Steel Flux Direct Flow Sanes Reactor Coolant Baffle-to- Former Assembly: PWR Vessel IV.B2-1 baffle and former plates SmuSteel and Neutron Change in Dimensions Internals (R-124) 3.1.1-33 A, 1 eStructural Flux Support

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 3/ Page 3 Table 3.1.2-3 REACTOR VESSEL INTERNALS Summary of Aging Management Evaluation Aging Effect Requiring Aging NUREG Table Intended Component Type Material Environment Management 1801 Vol. 3.X.1 Note Function Management Program 2 Item Item Direct PWR Vessel Flow Stainless Reactor Coolant Internals IV.B2-2 A, 1 Baffle-to- Former Assembly:

baffle and former plate Stainls and Neutron Cracking nea 123) 3.1.1-30 Structural Steel Flux Water Chemistry (R-123) A, 1 Support PWR Vessel A, I Baffle-to- Former Assembly: Structural Stainless and Neutron Cracking IV.B2-2 3.1.1-30 baffle-edge bolts Support Steel Flux Water Chemistry A, 1 Baffle-to- Former Assembly: Structural Stainless Reactor Coolant PWR Vessel IV.B2-6 baffle- e bortss Structr Stanles and Neutron Loss of Fracture Toughness Internals (R-128) 3.1.1-22 A, 1 baffle-edge bolts Support Steel Flux Baffle-to- Former Assembly: Structural Stainless Reactor Coolant PWR Vessel IV.B2-4 Support Steel and Neutron Change in Dimensions Internals (R-126) 3:1.1-33 A, 1 baffle-edge bolts FluxI Baffle-to- Former Assembly: Structural Stainless Reactor Coolant PWR Vessel IV.B2-5 Support Steel Flux Internals (R-129) baffle-edge bolts PWR Vessel A, I Baffle-to- Former Assembly: Structural Stainless and Neutron Cracking IV.B2-0 3.1.1-30 barrel-to-former bolts Support Steel Flux Water Chemistry (R125) 3 A, Fluxto WatrolantA, Baffle-to- Former Assembly: Structural Stainless Reactor Coolant PWR Vessel IV.B2-6 Support Steel Flux Internals (R-128) 3.1.1-22 A, I barrel-to-former bolts Baffle-to- Former Assembly: Structural Stainless Reactor Coolant PWR Vessel IV.B2-4 Baffle-to-Former bolts barrel-to-former tss l Structu Support Stanle Steel and Neutron Fu Change in Dimensions Internals (R-126) 3.1.1-33 A, 1 Flux eas Baffle-to- Former Assembly: Structural Stainless Reactor Coolant PWR Vessel IV.B2-5 Support Steel Flux Internals (R- 129) barrel-to-former bolts Bottom-mounted instrumentation system: bottom-mounted Structural Stainless Reatron PWR Vessel None None A, I instrumentation (BMI) column Support Steel Flux Internals bodies

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 3/ Page 4 Table 3.1.2-3 REACTOR VESSEL INTERNALS Summary of Aging Management Evaluation Intended Aging Effect Requiring Aging NUREG Table Component Type Function Material Environment Management Management 1801 Vol. 3.X.1 Note CmMntyFni i EProgram 2 Item Item Bottom-mounted instrumentation Reactor Coolant system: bottom-mounted Structural Stainless and Neutron Loss of Fracture Toughness Int el (R-1) 3.1.1-22 A, 1 instrumentation (BMI) column Support Steel FluxInternals (R-141) bodies Flux Control rod guide tube (CRGT) Structural Stainless Reactor Coolant PWR Vessel assemblies: CRGT guide plates Support Steeland Neutron Loss of Material Internals None None A, I (cards) Flux Control rod guide tube (CRGT) Coolant PWRVess1Reactor It el A assemblies: CRGT lower flange Surt Steel and Neutron Cracking Internals IV.B2-28 3.1.1-37 welds Support Steel Flux Water Chemistry (R-118) A,1 Control rod guide tube (CRGT) Stainless Reactor Coolant Structural Steel PWR Vessel IV.B2-22 assemblies: CRGT lower flange Support (includin and Neutron Loss of Fracture Toughness Internals (R-141) 3.1.1-22 A, I welds uppoS) ~CASS) cng Flux Stainless PWR Vessel Control rod guide tube (CRGT) Staile Reactor Coolant Internals IV.B2-28 A, I assemblies: guide tube support pins (split pins) Structural Support Steel; Nickel and FluxNeutron Cracking 1 8 (R-118) 3.1.1-37 A, 1 alloy Water Chemistry Control rod guide tube (CRGT) Stainless Reactor Coolant assemblies: guide tube support Structural Steel; and Neutron Loss of Material PWR Vessel None None A, I pins (split pins) Support Nickel Flux Internals alloy Direct PWR Vessel Core barrel assembly: upper core Flow Stainless Reactor Coolant Internals IV.B2-8 A, 1 barrel and lower core barrel Steel and Neutron Cracking (R-120) 3.1.1-30 circumferential (girth) welds Structural Flux Water Chemist A, 1 Support WaterChemistry Direct Core barrel assembly: upper core Flow Stainless Reactor Coolant PWR Vessel IV.B2-9 barrel and lower core barrel Steel and Neutron Loss of Fracture Toughness Internals (R-122) 3.1.1-22 A, I circumferential (girth) welds Structural Flux Support I I I I

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 3/ Page 5 Table 3.1.2-3 REACTOR VESSEL INTERNALS Summary of Aging Management Evaluation Intended Material Environment Aging Effect Requiring Aging NUREG Table Component Type Function Management Management Program 1801 Vol.

2 Item 3.X.1 Item Note Direct PWR Vessel Core barrel assembly: upper core Flow Stainless Reactor Coolant InteWrals IV.B2-8 A, 1 barrel and lower core barrel ames and Neutron Cracking (R- 120) 3.1.1-30 vertical (axial) welds Structural Flux Water Chemistry A, 1 Support Direct Core barrel assembly: upper core Flow Stal Reactor Coolant PWR Vessel IV.B2-9 barrel and lower core barrel Steel and Neutron Loss of Fracture Toughness Internals (R- 122) 3.1.1-22 A, I vertical (axial) welds Structural Flux Support Direct Core barrel assembly: core barrel Flow Stainless Reactor CoolantVessel Steel and Neutron Loss of Material Interals V None None A, flange Structural Flux Support Direct PWR Vessel FlwReactor Coolant A, I Core barrel assembly: core barrel otenozewlsand Flow Stainless ReutronNeutron Cracking Internals IV.B2-8

'R2~3.1.1-30 A,11 outlet nozzle welds Structural Steel Flux Water Chemistry A, I Support Direct Core barrel assembly: core barrel Flow Stainless Reactor Coolant PWR Vessel IV.B2-9 Structural Steel andFlux tron Loss of Fracture Toughness Internals (R-122) 3.1.1-22 A, outlet nozzle welds Support Direct PWR Vessel Flow Reactor Coolant Intess A, 1 Core barrel assembly: lower core Stainless and Neutron Cracking Internals IV.B2-8 3.1.1-30 barrel flange weld Structural Steel Flux Water Chemistry A, 1 Support Direct PWR Vessel Flow Reactor Coolant Intess A, 1 Core barrel assembly: upper core Stainless and Neutron Cracking Internals IV.B2-8 3.1.1-30 barrel flange weld Structural Steel Flux Water Chemistry A, 1 Support

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 3/ Page 6 Table 3.1.2-3 REACTOR VESSEL INTERNALS Summary of Aging Management Evaluation Environment Aging Effect Requiring Aging NUREG Table Intended Material Management 1801 Vol. 3.X.1 Note Component Type Function Management Program 2 Item Item Lower internals assembly: clevis Structural Nickel Reactor Coolant PWR Vessel Support Alloy and Neutron Loss of Material Internals None None A, I insert bolts or screws Flux Reactor Coolant Lower internals assembly: clevis Structural Nickel and Neutron Loss of Preload PWR Vessel IV.B2-14 3.1.1-27 A, I insert bolts or screws Support Alloy Flux Internals (R-137)

RatrCoatPWR Vessel A, I Lower internals assembly: clevis Structural Nickel Reactor Coolant Internals IV.B2-16 insert bolts or screws Support Alloy and Neutron Flux Cracking Water eChemistry s ~ (R- 13i 3.1.1-37 A, 1 Direct PWR Vessel Flow Stainless Reactor Coolant Internals IV.B2-20 A, 1 Lower internals assembly: lower and Neutron Cracking ) 3.1.1-30 core plate Structural Steel Flux Water Chemistry A, 1 Support Direct Lower internals assembly: lower Flow Stainless Reactor Coolant PWR Vessel IV.B2-18 and Neutron Loss of Fracture Toughness Internals (R-132) 3.1.1-22 A, 1 core plate Structural Steel Flux Support Reactor Coolant PWR Vessel A, 1 Lower support assembly: lower Structural CASS and Neutron Cracking Internals IV.132-24 3.1.1-30 support column bodies (cast) Support Flux Water Chemistry (R138) A, 1 Lower support assembly: lower Structural Reactor Coolant PWR Vessel Support CASS and Neutron Loss of Fracture Toughness Internals None None support column bodies (cast)

Flux Reactor Coolant PWR Vessel A, 1 Lower support assembly: lower Structural Stainless and Neutron Cracking Internals IV.B2-24 3.1.1-30 support forging or casting Support Steel Flux Water Chemistry (R-138) A, I Lower support assembly: lower Structural Stainless Reactor Coolant PWR Vessel IV.B2-17 3.1.1-22 A, forgingasti Lowersupport w Strutu Stanles and Neutron Loss of Fracture Toughness Internals (R-135) support forging or casting Support Steel Flux II

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 3/ Page 7 Table 3.1.2-3 REACTOR VESSEL INTERNALS Summary of Aging Management Evaluation Intended Aging Effect Requiring Aging NUREG Table Component Type Fut Management Management 1801 Vol. 3.X.1 Note Program 2 Item Item Lower support assembly: lower Structural Stainless Reactor Coolant PWR Vessel A, I and Neutron Cracking Internals IV.B2-24 3.1.1-30 support column bodies (non-cast) Support Steel Flux Water Chemistry (R-138) A, I Lower support assembly: lower Structural Stainless Reactor Coolant PWR Vessel IV.B2-9 support column bodies (non-cast) Support Steel and Neutron Loss of Fracture Toughness Internals (R-122) 3.1.1-22 A, I Flux Lower support assembly: lower Structural Stainless Reactor Coolant and Neutron PWR Vessel A, I Cracking Intemals IV.B2-16 3.1.1-30 support column bolts Support Steel Flux Water Chemistry (R-133) A, I Lower support assembly: lower Structural Stainless Reactor Coolant PWR Vessel IV.B2-17 support column bolts Support Steel and Neutron Loss of Fracture Toughness Internals (R-135) 3.1.1-22 A, I Flux Lower support assembly: lower Structural Stainless Reactor Coolant PWR Vessel IV.B2-25 supportrsupport cu colunm mnbolys bolts Structu Support Stanles Steel and Neutron Flux Loss of Preload Internals (R-136) 3.1.1-27 A, 1 Direct Stainless Reactor vessel internal Steel; IV.B2-31 components Nickel and Neutron Cumulative fatigue damage TLAA (R-53) 3.1.1-5 A, 1 Structural Support All Alloy Flux Direct Stainless Reactor vessel internal Flow Steel; 8 ,IV.B2-32 components Nickel and Neutron Loss of Material Water Chemistry (RP-24) 3.1.1-83 A, I sStructural Alloy Flux Support Direct ASME Section Xl Reactor vessel inteals: ASMEti Stainless Section X1, Examination Flow Reactor Coolant Inservice Category B-N-3 core support Steel; and Neutron Cracking Inspection, IV.B2-26 None A, 1 structure components (not already Structural Nickel Flux Subsections IWB, (R-142) identified as "Existing Programs" Strt Alloy IWC, and IWD components in MRP-227-A) Support_1WC, andIWD

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 3/ Page 8 Table 3.1.2-3 REACTOR VESSEL INTERNALS Summary of Aging Management Evaluation Aging Effect Requiring Aging NUREG Table Intended Component Type Function Material Environment Management Management 1801 Vol. 3.X.I Note FunctionManagementProgram 2 Item Item Reactor vessel internals: ASME Direct StainlessASME Section X Section XI, Examination Category B-N-3 core support Flow Steel; Reactor and Coolant Neutron Loss of Material Inspection, Inservice (R-12) None A, 1 Nickel Flux Subsections, (R-142)IoB, structure components (not already Structural Alloy FlC, and IWD identified as "Existing Programs" Support IWC, and IWD components in MRP-227-A)

Thermal shield assembly: thermal Structural Stainless Reactor Coolant PWR Vessel Support Steel and Neutron Cracking Internals shield flexures ReatorCooatFlux Thermal shield assembly: thermal Structural Stainless Reactor Coolant PWR Vessel Support Steel and Neutron Loss of Material Internals shield flexures

_________________________Flux No additional aging management Direct for reactor internal "No StaisFlow Reactor Coolant Additional Measures" Reactor internal "No Additional steel; and Neutron components unless required by IntePrals None None A, Measures" components Structural Nickel Flux ASME Section XI, Examination Support alloy Category B-N-3 or relevant operating experience exists Direct Upper Internals Assembly; upper Flow Stainless Reactor Coolant PWR Vessel None None A, 1 core plate coepaeSel Structural Steel FluxNeutron and Cracking Internals Support Direct Upper InteFals Assembly; upper Stainless Reactor Coolant PWR Vessel Uppe Inte Steel and Neutron Loss of Material Internals None None A, I core plate Structural Flux Support Reactor Coolant PWR Vessel A,1 Upper Internals Assembly: upper Structural Stainless and Neutron Cracking Internals IV.B2-42 3.1.1-30 support ring or skirt Support Steel Flux Water Chemistry (R-106) A, I

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 3/ Page 9 Standard Notes:

A Consistent with NUREG- 1801 item for component, material, environment, and aging effect. AMP is consistent with NUREG- 1801 AMP.

B Consistent with NUREG-1801 item for component, material, environment, and aging effect. AMP takes some exceptions to NUREG- 1801 AMP.

C Component is different, but consistent with NUREG-1801 item for material, environment, and aging effect. AMP is consistent with NUREG- 1801 AMP.

D Component is different, but consistent with NUREG- 1801 item for material, environment, and aging effect. AMP takes some exceptions to NUREG- 1801 AMP E Consistent with NUREG- 1801 for material, environment and aging effect, but a different aging management program is credited or NUREG- 1801 identifies a plant-specific aging management program F Material not in NUREG- 1801 for this component.

G Environment not in NUREG-1801 for this component and material.

H Aging effect not in NUREG-1801 for this component, material and environment combination.

I Aging effect in NUREG- 1801 for this component, material and environment combination is not applicable.

J Neither the component nor the material and environment combination is evaluated in NUREG- 1801.

Plant Specific Notes:

1 Consistent with NUREG-1801 as modified by LR-ISG-2011-04.

Enclosure 4 to SBK-L-14089 LRA Appendix A - Final Safety Report Supplement Table A.3, License Renewal Commitment List Updated to Reflect Changes to Date

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 4/ Page 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.

A Inspection

.. plan to be ___,

_-;________.

.. sub.itted

____÷*A to

÷ An inspection plan for-Reaetor: Vessel internals will be submitted for-NRC NRC not later than 2 years afe review and approval, receipt of the renewed license or-not PWR Vessel Internals Provide confirmation and acceptability of the implementation of MRP-227- A.2.1.7 less than 24 months prior to the A by addressingthe plant-specificApplicant/LicenseeAction Items outlined period of extended operation,-

in section 4.2 of the NRC SER. whichever comes first.

NextEra Energy Seabrook will pro vide a subinittalschedule by October 15, 2014.

Closed-Cycle Cooling Enhance the program to include visual inspection for cracking, loss 6f A.2.1.12 Prior to the period of extended

2. Water material and fouling when the in-scope systems are opened for maintenance. operation.

Inspection of Overhead Heavy Load and Light Enhance the program to monitor general corrosion on the crane and trolley Prior to the period of extended Refuelandling structural components and the effects of wear on the rails in the rail system. A.2.1.13 operation.

Refueling) Handling Systems Inspection of Overhead Heavy Load and Light Prior to the period of extended

4. Load (Related to Enhance the program to list additional cranes for monitoring. A.2.1.13 prratote Refueling) Handling operation.

Systems Compressed Air Enhance the program to include an annual air quality test requirement for the Prior to the period of extended

5. Monitoring Diesel Generator compressed air sub system. A.2.1.14 operation.

Enhance the program to perform visual inspection of penetration seals by a A.2.1.15 Prior to the period of extended

6. Fire Protection fire protection qualified inspector. operation.

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 4/ Page 3 Enhance the program to add inspection requirements such as spalling, and Prior to the period of extended

7. Fire Protection loss of material caused by freeze-thaw, chemical attack, and reaction with A.2.1.15 operation.

aggregates by qualified inspector.

Enhance the program to include the performance of visual inspection of fire- Prior to the period of extended

8. Fire Protection rated doors by a fire protection qualified inspector. A.2.1.15 operation.

Enhance the program to include NFPA 25 (2011 Edition) guidance for

9. Fire Water System "where sprinklers have been in place for 50 years, they shall be replaced or A.2.1.16 Prior to the period of extended representative samples from one or more sample areas shall be submitted to a operation.

recognized testing laboratory for field service testing".

Enhance the program to include the performance of periodic flow testing of Prior to the period of extended

10. Fire Water System the fire water system in accordance with the guidance of NFPA 25 (2011 A.2.1.16 operation.

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 Within ten years prior to the period

11. Fire Water System examinations are performed. These inspections will be documented and A.2.1.16 of extended operation.

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 be performed within ten years prior to the period of extended operation.

Enhance the program to include components and aging effects required by Pr.i to .the period of extended

12. Aboveground Steel Tanks the Aboveground Steel Tanks and to perform visual, surface, and volumetric A.2.1.17 .........

examinations of the outside and inside surfaces for managing the aging Within 10 years prior to the period effects of loss of material and cracking. of extended operation.

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 4/ Page 4 Enhance the program to perform exterior inspection of the fire water storage Aboveground Steel Tanrs tanks annually for signs of degradation and include an ultrasonic inspection A Within ten years prior to the period

13. Fire Water System and evaluation of the internal bottom surface of the two Fire Protection Water A.2.1.16 of extended operation.

Storage Tanks per the guidance provided in NFPA 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 Prior to the period of extended

14. Fuel Oil Chemistry storage tank and 2) periodically sample stored fuel in the Auxiliary Boiler A.2.1.18 operation.

fuel oil storage tank.

Enhance the program to add requirements to check for the presence of water Prior to the period of extended

15. Fuel Oil Chemistry in the Auxiliary Boiler fuel oil storage tank at least once per quarter and to A.2.1.18 operation.

remove water as necessary.

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

16. Fuel Oil Chemistry diesel fire pump fuel oil day tanks on a frequency of at least once every ten A.2.1.18 operation.

years.

Enhance the program to require ultrasonic thickness measurement of the tank bottom during the 10-year draining, cleaning and inspection of the Diesel Prior to the period of extended

17. Fuel Oil Chemistry Generator fuel oil storage tanks, Diesel Generator fuel oil day tanks, diesel A.2.1.18 operation.

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 capsules, unless A.2.1.19 Prior to the period of extended

18. Surveillance discarded before August 31, 2000, are placed in storage. operation.

Enhance the program to specify that if plant operations exceed the limitations Reactor Vessel or bounds defined by the Reactor Vessel Surveillance Program, such as Prior to the period of extended

19. Surveillance operating at a lower cold leg temperature or higher fluence, the impact of A.2.1.19 operation.

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

20. Surveillance meets the schedule requirements of 10 CFR 50 Appendix H and ASTM A.2.1.19 operation.

El 85-82 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.

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 4/ Page 5 Reactor Vessel Enhance the program to ensure that any capsule removed, without the intent Prior to the period of extended

21. Surveillance to test it, is stored in a manner which maintains it in a condition which would A.2.1.19 operation.

permit its future use, including during the period of extended operation.

Within ten years prior to the period

22. One-Time Inspection Implement the One Time Inspection Program. A.2. 1.20 of extended operation.

Implement the Selective Leaching of Materials Program. The program will Selective Leaching of include a one-time inspection of selected components where selective Within five years prior to the period

23. Materials leaching has not been identified and periodic inspections of selected A.2.1.21 of extended operation.

components where selective leaching has been identified.

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

24. Inspection the period of extended operation
25. AOne-Time Inspection of Implement the One-Time Inspection of ASME Code Class 1 Small Bore- A.2.1.23 Within ten years prior to the period ASME Code Class I Piping Program. of extended operation.

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

26. Monitoring inspection frequency, the inspections of opportunity for possible corroesion A.2.1.24 operation.

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

Inspection of Internal Surfaces in Miscellaneous Implement the Inspection of Internal Surfaces in Miscellaneous Piping and Prior to the period of extended

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

Components Enhance the program to add required equipment, lube oil analysis required, A.2.1.26 Prior to the period of extended

28. sampling frequency, and periodic oil changes. operation.

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

29. Lubricating Oil Analysis collection tanks. operation.

Enhance the program to require the performance of a one-time ultrasonic Prior to the period of extended

30. Lubricating Oil Analysis thickness measurement of the lower portion of the Reactor Coolant pump oil A.2.1.26 operation.

collection tanks prior to the period of extended operation.

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 4/ Page 6 ASME Section XI, Prior to the period of extended

31. Subsection IWL Enhance procedure to include the definition of"Responsible Engineer". A.2.1.28 operation.

Structures Monitoring Enhance procedure to add the aging effects, additional locations, inspection A.2.1.31 Prior to the period of extended

32. Program frequency and ultrasonic test requirements. operation.

Structures Monitoring Enhance procedure to include inspection of opportunity when planning A.2.1.31 Prior to the period of extended 3 Program excavation work that would expose inaccessible concrete. operation.

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

34. Environmental 50.49 Environmental Qualification Requirements program. operation.

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

35. Environmental 50.49 Environmental Qualification Requirements Used in Instrumentation A.2.1.33 operation.

Qualification Circuits program.

Requirements Used in Instrumentation Circuits Inaccessible Power Cables Not Subject to 10 Implement the Inaccessible Power Cables Not Subject to 10 CFR 50.49 Prior to the period of extended

36. CFR 50.49 Environmental Environmental Qualification Requirements program. operation.

Qualification Requirements Prior to the period of extended Implement the Metal Enclosed Bus program. A.2.1.35 operaton.

37. Metal Enclosed Bus operation.

Prior to the period of extended operaton.

A.2.1.36

38. Fuse Holders Implement the Fuse Holders program. operation.

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 4/ Page 7 Electrical Cable Connections Not Subject to 10 CFR 50.49 Implement the Electrical Cable Connections Not Subject to 10 CFR 50.49 A.2.1.37 Prior to the period of extended Environmental Environmental Qualification Requirements program. operation.

Qualification Requirements Prior to the period of extended Proraton.

Implement the 345 KV SF6 Bus program. A.2.2.1

40. 345 KV SF6 Bus operation.

Metal Fatigue of Reactor Enhance the program to include additional transients beyond those defined in Prior to the period of extended

41. Coolant Pressure the Technical Specifications and UFSAR. A.2.3.1 operation.

Boundary MtlFatigue of ReactorPrototepidofxend Metal Enhance the program to implement a software program, to count transients to Prior to the period of extended

42. Coolant Pressure monitor cumulative usage on selected components. operation.

Boundary Pressure -Temperature The updated analyses will be Limits, including Low submitted at the appropriate time to LiTmpertsinclureng Lo Seabrook Station will submit updates to the P-T curves and LTOP limits to A44m ith 10 CFR 50pAppendix

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

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 4/ Page 8 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 water enviromnent. This includes applying the appropriate Fen factors to valid Environmentally-Assisted At least two years prior to entering

44. CUFs determined from an existing fatigue analysis valid for the period of A.2.4.2.3 Fatigue Analyses (TLAA) 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 SBK-L-14089 / Enclosure 4/ Page 9

45. Number Not Used Protective Coating Enhance the program by designating and qualifying an Inspector Coordinator Prior to the period of extended
46. Monitoring and and an Inspection Results Evaluator. A.2..38 operation.

Maintenance Protective CoatingA2.38 Enhance the program by including, "Instruments and Equipment needed for oeain

47. Monitoring and inspection may include, but not be limited to, flashlight, spotlights, marker A.2.1.38 Prior to the period of extended Maintenance pen, mirror, measuring tape, magnifier, binoculars, camera with or without operation.

wide angle lens, and self sealing polyethylene sample bags."

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

48. Monitoring and A.2.1.38 operation.

Maintenance reports.

Protective Coating Enhance the program to require that the inspection report is to be evaluated

49. Monitoring and by the responsible evaluation personnel, who is to prepare a summary of A.2.1.38 oratote Maintenance findings and recommendations for future surveillance or repair. operation.

Within the next two refueling outages, OR15 or ORI6, andmor tha our ated at i r o Of ASME Section XI, Perform UT testing of the containment liner plate in the vicinity of the

50. Subsection IWE moisture barrier for loss of material. five refueling outages.
51. Number Not Used Section Implement measures to maintain the exterior surface of the Containment -A.2.1.28 Ongoing
52. ASME Subsection IWLXI, Structure, from elevation -30 feet to +20 feet, in a dewatered state.

Reactor Head Closure Replace the spare reactor head closure stud(s) manufactured from the bar that A.2.1.3 Prior to the period of extended Studs has a yield strength > 150 ksi with ones that do not exceed 150 ksi. operation.

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 4/ Page 10 NextEra will address the potential for cracking of the primary to secondary pressure boundary due to PWSCC of tube-to-tubesheet welds using one of the following two options:

1) Perform a one-time inspection of a representative sample of tube-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 Steam Generator Tube
54. routine tube-to-tubesheet weld inspections for the remaining life of the steam A.2. 1.10 Complete Integrity 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 of tube-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.

Steam Generator Tube Seabrook will perform an inspection of each steam generator to assess the A.2.1.10 Within five years prior to entering 5 Integrity condition of the divider plate assembly. the period of extended operation.

Closed-Cycle Cooling Revise the station program documents to reflect the EPRI Guideline A.2.1.12 Prior to entering the period of

56. Water System operating ranges and Action Level values for hydrazine and sulfates. extended operation.

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

57. Water System operating ranges and Action Level values for Diesel Generator Cooling A.2.1.12 extended operation.

Water Jacket pH.

Update Technical Requirement Program 5.1, (Diesel Fuel Oil Testing Prior to the period of extended

58. Fuel Oil Chemistry Program) ASTM standards to ASTM D2709-96 and ASTM D4057-95 A.2.1.18 operation.

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

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

60. Inspection piping with a pipe-within-pipe configuration with leak detection capability. A.2.1.22 extended operation.

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 4/ Page 11 Compressed Air Replace the flexible hoses associated with the Diesel Generator air Within ten years prior to entering 61 Monitoring Program compressors on a frequency of every 10 years. A.2.1.14the period of extended operation.

Enhance the program to include a statement that sampling frequencies are A21.2 Prior to the period of extended

62. Water Chemistry increased when chemistry action levels are exceeded. A. 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 Prior to the period of extended

63. Flow Induced Erosion increase in the CVCS Charging Pump mini flow above the acceptance N/A operation.

criteria 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- Prior to entering the period of

64. Buried Piping and Tanks cathodically protected steel pipe within the scope of this program. If the A.2.1.22 extended operation.

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

Implement measures to ensure that the movable incore detectors are not Prior to entering the period of

65. Flux Thimble Tube returned to service during the period of extended operation. N/A extended operation.
66. Number Not Used Perform one shallow core bore in an area that was continuously wetted from Structures Monitoring borated water to be examined for concrete degradation and also expose rebar to any to degradation such as loss of material. removed core will also A.2.1.31 The degradation No later than December 31, 2015.
67. Program be detect subjected petrographic examination for concrete due to ASR per ASTM Standard Practice C856.

Structures Monitoring Perform sampling at the leakoff collection points for chlorides, sulfates, pH A.2.1.31 Quarterly Preventive Maintenance

68. Program and iron once every three months. Activity Implemented Open-Cycle Cooling Replace the Diesel Generator Heat Exchanger Plastisol PVC lined Service Prior to the period of extended
69. Water System Water piping with piping fabricated from AL6XN material. A.2.1.11 operation.

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

70. Water System eliminated or whether this remains an issue in the primary component A.2.1.12 of extended operation.

cooling water system.

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 4/ Page 12 Alkali-Silica Reaction Implement the Alkali-Silica Reaction (ASR) Monitoring Program. Testing Prior to entering the period of

71. (ASR) Monitoring will be performed to confirm that parameters being monitored and acceptance A.2.1.31A extended operation.

Program criteria used are appropriate to manage the effects of ASR.

Flow-Accelerated Enhance the program to include management of wall thinning caused by A.2.1.8 Prior to entering the period of

72. Corrosion mechanisms other than FAC. extended operation.

Inspection of Internal Enhance the program to include performance of focused examinations to Surfaces in Miscellaneous provide a representative sample of 20%, or a maximum of 25, of each A.2.1.25 Prior to entering the period of Piping and Ducting identified material, environment, and aging effect combinations during each extenterin .

Components 10 year period in the period of extended operation. extended operation.

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 Within ten years prior to the period

74. Fire Water System and installed in the proper orientation (e.g., upright, pendant, or sidewall). A.2.1.16 of extended operation.

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 conduct an inspection of piping and branch line conditions every 5 years by opening a flushing connection at the end of one Within ten years prior to the period

75. Fire Water System main and by removing a sprinkler toward the end of one branch line for the A.2.1.16 of extended operation.

purpose of inspecting for the presence of foreign organic and inorganic material per the guidance provided in NFPA 25 (2011 Edition).

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

76. Fire Water System
  • main drain tests A.2.1.16 Wh of tenyea extended priori operation.

0 deluge valve trip tests 0 fire water storage tank exterior surface inspections

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 4/ Page 13 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).

77. The requirement that if there is a 10 percent reduction in full flow A.2.1.16 Within ten years prior to the period pressure when compared to the original acceptance tests or previously of extended operation.

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

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

78. External Surfaces Enhance the program to include periodic inspections of in-scope insulated A.2.1.8 Prior to the period of extended Monitoring components for possible corrosion under insulation. operation.

Enhance the program to include visual inspection of Service Level III

79. Open-Cycle Water SystemCooling (augmented) internal coatings for loss of coating integrity. A.2.1.11 Witended of prion.

extended operation.

Enhance the program to include visual inspection of Service Level IIl Within 10 years prior to the period

80. Fire Water System (augmented) internal coatings for loss of coating integrity. A.2.1.16 of extended operation.

Enhance the program to include visual inspection of Service Level III Within 10 years prior to the period

81. Fuel Oil Chemistry (augmented) internal coatings for loss of coating integrity. A.2.1.18 of extended operation.

Inspection of Internal Enhance the program to include visual inspection of Service Level III

82. Surfaces in Miscellaneous (augmented) internal coatings for loss of coating integrity. A.2.1.25 Within 10 years prior to the period Piping and Ducting of extended operation.

Components

United States Nuclear Regulatory Commission SBK-L-14089 / Enclosure 4/ Page 14 Install instrumentation in representative sample areas of structures to monitor expansion due to alkali-silica reaction in the out-of-plane direction. Evaluate instrument and pin expansion data under the Operating Experience Element Alkali-Silica Reaction of the Alkali-Silica Reaction Monitoring Program to determine whether there Prior to the period of extended

83. Monitoring is a need to enhance the program to monitor expansion in the out-of-plane A.2.1.31 A operation.

direction. If the evaluation concludes that out-of-plane monitoring is necessary, establish acceptance criteria and monitoring frequencies for expansion in the out-of-plane direction using the instrument and pin expansion data.

ASME Section XI, Evaluate the acceptability of inaccessible areas for structures within the scope Prior to the period of extended

84. A.2.1.28 Subsection IWL of ASME Section XI, Subsection IWL Program. operation.
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