SBK-L-15073, Responses to Applicant/Licensee Action Items for the Inspection and Evaluation Guidelines for Pressurized Water Reactor Vessel Internals (MRP-227-A)

Responses to Applicant/Licensee Action Items for the Inspection and Evaluation Guidelines for Pressurized Water Reactor Vessel Internals (MRP-227-A)

ML15149A279
Person / Time
Site:	Seabrook
Issue date:	05/26/2015
From:	Dean Curtland NextEra Energy Seabrook
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML15149A278	List: ML15149A280 ML15149A282 ML15149A283 SBK-L-15073, Responses to Applicant/Licensee Action Items for the Inspection and Evaluation Guidelines for Pressurized Water Reactor Vessel Internals (MRP-227-A)
References
MRP-227-A, SBK-L-15073, TAC ME4028
Download: ML15149A279 (72)

Text

Enclosure 10 to this Letter Contains Proprietary Information Withhold Enclosure 10 from Public Disclosure in Accordance with 10 CFR 2.390 NExTerao ENERGY01

_,ýýýSEABROKý May 26, 2015 10 CFR 54 SBK-L- 15073 Docket No. 50-443 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Seabrook Station Responses to Applicant/Licensee Action Items for the Inspection and Evaluation Guidelines for Pressurized Water Reactor Vessel Internals (MRP-227-A)

References:

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

2. NextEra Energy Seabrook, LLC letter SBK-L-14089, "Response to Request for Additional Information Related to the Review of The Seabrook Station License Renewal Application- Set 21 (Tag No. ME4028)," June 24, 2014. (Accession Number ML14177A502)

3. NextEra Energy Seabrook, LLC letter SBK-L- 14212, "Update to the PWR Vessel Internals Program," November 21, 2014.

4. LR-ISG-2011-04, "Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors," May 28, 2013.

5. EPRI Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A), Technical Report 1022863, December 2011.

6. Revision I to the NRC Safety Evaluation Report for EPRI Materials Reliability Program:

Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A),

December 16, 2011.

In Reference 1, NextEra Energy Seabrook, LLC (NextEra Energy Seabrook) submitted an application for a renewed facility operating license for Seabrook Station Unit 1 in accordance with the Code of Federal Regulations, Title 10, Parts 50, 51, and 54.

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

U.S. Nuclear Regulatory Commission SBK-L-15073/Page 2 In Reference 2, NextEra Energy submitted a revised PWR Vessel Internals Program using the guidance provided in LR-ISG-2011-04 (Ref. 4) and MRP-227-A (Ref. 5).

In Reference 3-, NextEra Energy provided a due date of May 31, 2015 for providing the 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 Safety Evaluation Report for MRP-227-A (Ref. 6). of this supplement letter provides the NextEra Energy Seabrook non-proprietary responses to Applicant/Licensee Action Items for the Inspection and Evaluation Guidelines for Pressurized Water Reactor Vessel Internals (MRP-227-A). provides the revised NextEra Energy Seabrook PWR Vessel Internals Program.

Tables 1 through 4 of the PWR Vessel Internals Program provide the NextEra Energy Seabrook PWR Vessel Internals Inspection Plan. It should be noted that the PWR Vessel Internals Program that was submitted in SBK-L-14089, dated June 24, 2014 (Ref. 2) is superseded by this supplement letter. provides the changes to the UFSAR Supplement (Section A.2.1.7) and changes to the commitments related to the PWR Vessel Internals Program. provides the revised AMR items for the Reactor Vessel Internals. It should be noted that the AMR line items that were submitted in SBK-L- 14089, dated June 24, 2014 (Ref. 2) are superseded by this supplement letter. provides the revised LRA Appendix A - Updated Final Safety Analysis Report Supplement Table A.3, License Renewal Commitment List. Commitment #1 has been completed and a new Commitment #90 has been added in this Supplement letter.

Enclosures 6 and 7 provide Non-Proprietary Class 3 Westinghouse Reports prepared in support of NextEra Energy Seabrook response to Applicant/Licensee Action Items 1, 2, and 7. contains the Non-Proprietary Appendix A of Westinghouse Report (PWROG-15042-P, Revision 0) - NextEra Energy, Seabrook Unit 1 Summary Report for the Cold Work Assessment - Appendix A

U.S. Nuclear Regulatory Commission SBK-L-15073/Page 3 0 provides information proprietary to Westinghouse Electric Company LLC, which was prepared in support of NextEra Energy Seabrook response to Applicant/Licensee Action Item 1 and is supported by an affidavit in Enclosure 9 signed by Westinghouse, the owner of the information. The affidavit sets forth the basis on which the information may be withheld from public disclosure and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390. Accordingly, NextEra requests that the information that is proprietary to Westinghouse (Enclosure 10) be withheld from public disclosure in accordance with 10 CFR 2.390. Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Westinghouse affidavit should reference CAW-15-4190 and should be addressed to J.A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, Suite 428, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.

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.

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

Executed on May2tb, 2015.

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

U.S. Nuclear Regulatory Commission SBK-L-15073/Page 4

Enclosures:

- NextEra Energy Seabrook Responses to Applicant/Licensee Action Items for the Inspection and Evaluation Guidelines for Pressurized Water Reactor Vessel Internals (MRP-227-A) - Revised NextEra Energy Seabrook PWR Vessel Internals Program Including the PWR Vessel Internals Inspection Plan - Changes to the UFSAR Supplement (Section A.2.1.7) and Changes to the Commitments related to the PWR Vessel Internals Program - Revised AMR Items for the Reactor Vessel Internals (Revised LRA Table 3.1.2-3) - LRA Appendix A - Updated Final Safety Analysis Report Supplement Table A.3, License Renewal Commitment List Updated to Reflect Changes to Date - Non-Proprietary Class 3 Westinghouse Report (PWROG-14083-NP, Revision 1) -

NextEra Energy, Seabrook Unit 1 Summary Report for Applicant/Licensee Action Items 1, 2, and 7 - Non-Proprietary Class 3 Westinghouse Report (PWROG-15023-NP, Revision 1) -

Seabrook Station Unit 1 Summary Report for the Fuel Design/Fuel Management Assessments to Demonstrate MRP-227-A Applicability - Appendix A of Westinghouse Report (PWROG-15042-P, Revision 0) - NextEra Energy, Seabrook Unit 1 Summary Report for the Cold Work Assessment (Non-Proprietary) - Application for Withholding Proprietary Information from Public Disclosure and Affidavit 0 - Proprietary Class 2 Westinghouse Report (PWROG-1 5023-P, Revision 1) -

NextEra Energy Seabrook Unit I Summary Report for the Fuel Design/Fuel Management Assessments to Demonstrate MRP-227-A Applicability cc: D. H. Dorman 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

U.S. Nuclear Regulatory Commission SBK-L-15073/Page 5 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-15073 NextEra Energy Seabrook Responses to Applicant/Licensee Action Items For the Inspection and Evaluation Guidelines for Pressurized Water Reactor Vessel Internals (MRP-227-A)

(Non-Proprietary)

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure I/Page 2 NextEra Energy Seabrook Responses to Applicant/Licensee Action Items for the Inspection and Evaluation Guidelines for Pressurized Water Reactor Vessel Internals (MRP-227-A)

Applicant/Licensee Action Item 1: Applicability of FMECA and Functionality Analysis Assumptions (NRC SE Section 4.2.1)

Applicability of FMECA and Functionality Analysis Assumptions text from NRC SER in MRP-227-A (Ref. 5) states:

"As addressedin Section 3.2.5.1 of this SE, each applicant/licenseeis responsiblefor assessing its plant's design and operatinghistory and demonstratingthat the approved version qfMRP-227 is applicable to the facility. Each applicant/licenseeshall refer, in particular,to the assumptions regardingplant design and operatinghistory made in the FMECA andfunctionality analysesfor reactorsof their design (i.e., Westinghouse, CE, or B& W) which support MRP-22 7 and describe the process usedfor determiningplant-specific differences in the design of their R VI components or plant operatingconditions,which result in different component inspection categories. The applicant/licenseeshall submit this evaluationfor NRC review and approval as part of its applicationto implement the approved version of MRP-227. This is Applicant/Licensee Action Item 1 ".

NextEra Energy Seabrook Response to Applicant/Licensee Action Item 1 MRP-227-A, Section 2.4, includes three bounding assumptions concerning its applicability to individual licensees: 1) Operation of 30 years or less with high-leakage core loading patterns (fresh fuel assemblies loaded in peripheral locations) followed by implementation of a low-leakage fuel strategy for the remaining 30 years of operation, 2) Base load operation (i.e.,

typically operates at fixed power levels and does not vary power on a calendar or load demand schedule), and 3) No design changes beyond those identified in general industry guidance or recommended by the original vendors. These bounding assumptions are directly related to Applicant/Licensee Action Item 1 and, as such, are addressed in the response to this Applicant/Licensee Action Item.

The process used to provide reasonable assurance that the RVI components at NextEra Energy Seabrook are reasonably represented by the generic industry program assumptions (with regard to neutron fluence, temperature, stress values, and materials used in the development of MRP-227-A) is:

1. Identification of typical Westinghouse designed pressurized water reactor (PWR) RVI components as described in MRP- 191, Table 4.4 (Ref. 1).

2. Identification of NextEra Energy Seabrook RVI components.

3. Comparison of the typical Westinghouse designed PWR RVI components to the NextEra Energy Seabrook RVI components:

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 1/Page 3

a. Confirmation that no additional items were identified by this comparison (primarily supports Applicant/Licensee Action Item 2).

b. Confirmation that the materials from MRP- 191, Table 4-4 are consistent with NextEra Energy Seabrook RVI component materials.

c. Confirmation that the design and fabrication of NextEra Energy Seabrook RVI components are the same as, or equivalent to, the typical Westinghouse designed PWR RVI components.

4. Confirmation that the NextEra Energy Seabrook operating history is consistent with the assumptions in MRP-227-A regarding core loading patterns and base load operation.

5. Confirmation that the NextEra Energy Seabrook RVI materials operated at temperatures within the original design basis parameters.

6. Determination of stress values based on design basis documents.

7. Confirmation that any changes to the NextEra Energy Seabrook RVI components do not impact the application of the MRP-227-A generic aging management strategy.

The NextEra Energy Seabrook RVI components are reasonably represented by the design and operating history assumptions regarding neutron fluence, temperature, materials, and stress values in the MRP-191 generic failure modes, effects, and criticality analysis (FMECA), and in the MRP-232 (Ref. 11) functionality analysis based on the following:

1. NextEra Energy Seabrook operating history is consistent with the assumptions in MRP-227-A with regard to neutron fluence and fuel management.

a. The FMECA and functionality analysis for MRP-227-A were based on the assumption of 30 years of operation with high-leakage core loading patterns, followed by 30 years of low-leakage core fuel management strategy. As stated in Reference 8, NextEra Energy Seabrook fuel management program changed from a high to a low-leakage core loading pattern prior to 30 years of operation. By operating with a low-leakage core design prior to 30 years, NextEra Energy Seabrook meets the fluence and fuel management assumptions in MRP-191 and requirements for MRP-227-A application.

b. NextEra Energy Seabrook has always operated as a base load unit. Therefore, NextEra Energy Seabrook satisfies the assumptions in MRP documents regarding operational parameters affecting fluence.

2, The NextEra Energy Seabrook reactor coolant system operates between Tcold and Thot (Ref.

9). TcoId is no lower than 537.7°F and Thot is no higher than 621.4°F (Ref. 9, Table 5.1-1).

The design temperature for the vessel is 650'F (Ref. 9, Table 5.3-1). Therefore, NextEra Energy Seabrook operating history is within original design basis parameters and is

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 1/Page 4 consistent with the assumptions used to develop the MRP-227-A aging management strategy with regard to temperature operational parameters.

3. As discussed below, the NextEra Energy Seabrook RVI components and materials are comparable to the typical Westinghouse designed PWR RVI components described in MRP-191, Table 4-4.

a. Components required to be in the NextEra Energy Seabrook aging management program are consistent with those contained in MRP-191, Table 4-4. No additional components were identified for NextEra Energy Seabrook by this comparison.

b. NextEra Energy Seabrook RVI component materials are consistent with, or equivalent to, those materials identified in MRP- 191, Table 4-4 for Westinghouse designed plants (Ref.

Table 2 attached to the end of this enclosure). Where differences exist, there is no impact on the NextEra Energy Seabrook PWR RVI aging management program or the component is already credited as being managed under an alternate NextEra Energy Seabrook aging management program.

c. Design and fabrication of NextEra Energy Seabrook PWR RVI components are the same as, or equivalent to, the typical Westinghouse designed PWR RVI components, with the exception that NextEra Energy Seabrook utilizes a double-concentric thimble tube design fabricated from wear resistant, seamless nickel alloy material (INCONEL 600).

4. NextEra Energy Seabrook Station is a single unit, 1295 net megawatts electric Westinghouse 4-loop pressurized water reactor with a turbine generator built by General Electric.

Commercial operation began in August 1990 with a design rated power of 3411 megawatts thermal (MWt). Two power uprates have been implemented since initial commercial operation. In Cycle 11 (2004), the rated thermal power was increased to 3587 MWt, and in Cycle 12 (2006), the rated thermal power was increased to 3648 MWt (Ref. 9).

The original Alloy X-750 guide tube assembly split pins were replaced with cold worked 316 SS in 2006. Therefore, modifications to the NextEra Energy Seabrook RVI made over the lifetime of the plant are those specifically directed by the original equipment manufacturer (OEM). The OEM has developed or evaluated design changes and satisfied assumptions for Applicant/Licensee Action item 1. The design has been maintained over the lifetime of the plant as specified by the OEM, operational parameters are compliant with MRP-227-A requirements with regard to fluence and temperature, and the components are consistent with those considered in MRP-191. The materials for the components are consistent with those considered in MRP- 191. Therefore, the NextEra Energy Seabrook RVI stress values are represented by the assumptions in MRP-191, MRP-227-A, and MRP-232 confirming the applicability of the generic FMECA.

Additionally, in Reference 15, the NRC staff indicated that information provided by the industry

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 1/ Page 5 to the NRC staff demonstrated that the MRP-227-A Inspection and Evaluation Guidelines are applicable for the range of conditions expected at the currently operating Westinghouse and Combustion Engineering designed plants in the United States. As a result of technical discussions with the NRC staff, the basis for a plant to respond to the NRC's Request for Additional Information (RAI) to demonstrate compliance with MRP-227-A for originally licensed and uprated conditions was determined to be satisfied with plant-specific responses to the following two questions (Ref. 15 and 16):

1. Does the plant have non-weld or bolting austenitic stainless steel (SS) components with 20 percent cold work or greater, and, if so, do the affected components have operating stresses greater than 30 ksi? (If both conditions are true, additional components may need to be screened in for stress corrosion cracking, SCC).

2. Does the plant have atypical fuel design or fuel management that could render the assumptions of MRP-227-A, regarding core loading/core design, non-representative for that plant?

NextEra Energy Seabrook Evaluation for Question 1 Westinghouse has evaluated the NextEra Energy Seabrook RVI components according to industry guideline provided in MRP 2013-025 (Ref. 17) and MRP-191 industry generic component listings and screening criteria [(including consideration of cold work as defined in MRP- 175 (Ref. 18), noting the requirements of subsection 3.2.3)]. In addition to consideration of the material fabrication, forming and finishing process, a general screening definition of "severe cold work" as a resulting reduction in wall thickness of 20 percent was applied as an evaluation limit. It was confirmed that all of NextEra Energy Seabrook components, as applicable for the design, are included directly in the MRP- 191 component lists, or have been evaluated accordingly.

The evaluation included a review of all plant modifications affecting reactor internals and the plant operating history. The components were procured according to the American Society for Testing and Materials (ASTM) International or American Society of Mechanical Engineers (ASME) material specifications that were called out on the original plant construction drawings.

Material and component procurement was through applicable quality-controlled protocols.

Therefore, material identification based on the material call-outs and notes in the component drawings was an efficient and reasonable approach to identify the materials of construction for the RVI components at NextEra Energy Seabrook.

Based on the specifications called out on the NextEra Energy Seabrook component drawings, the RVI components are binned into the five material categories identified in MRP 2013-025.

Cold work categories based on MRP 2013-025 include:

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 1/ Page 6

" Cast austenitic stainless steel (CASS) (Category 1)

• Hot-formed austenitic stainless steel (Category 2)

• Annealed austenitic stainless steel (Category 3)

" Fasteners austenitic stainless steel (Category 4)

" Cold-formed austenitic stainless steel without subsequent solution annealing (Category 5)

The potential for cold work is directly controlled by the materials specifications. Essentially all of the components that are binned (based on their specified materials) as Categories 1, 2 and 3 are non-cold worked; therefore, they have less than 20 percent cold work according to the NRC criterion. Similarly, any component binned under Category 5 has the potential to contain greater than 20 percent cold work. Category 4 materials are fasteners that may have been intentionally strain-hardened.

During the fabrication of fasteners, the strain hardening was typically intentionally restricted to less than 20 percent. These restrictions, if present, were noted on engineering drawings. A restriction or limitation on the material yield stress (e.g., a maximum of 90 ksi) would indicate that the material cold work would be limited to be less than 20 percent. In the absence of a restriction on the maximum yield stress of strain-hardened material, a conservative approach has been taken to assume the potential for greater than 20 percent cold work.

Where multiple options existed for a component or assembly, the bounding condition of cold work was taken as the option that had the greater potential to include greater than 20 percent cold work. This option was then employed in the assessment of the component, and was selected for the purposes of the assessment. In some instances, sequential fabrication would appear to mitigate any potential for cold work; however, since the historical record was not detailed the potential is noted, but a conservative approach was selected for this assessment.

The evaluation, performed consistently with the industry guideline provided in MRP 2013-025, concluded that the reactor internals Category 1, 2 and 3 (non-bolting) components at NextEra Energy Seabrook contain no cold work greater than 20 percent as a result of material specification and controlled fabrication construction. No NextEra Energy Seabrook components were binned as Category 5. Therefore, the only materials with the potential for greater than 20 percent cold work NextEra Energy Seabrook were strain-hardened fasteners binned as Category 4 components. For some Category 4 components, the material drawing notes and Westinghouse purchasing specifications that were employed in addition to ASME and ASTM specifications for parts purchase were found to limit the strength of the employed materials such that the use of greater than 20 percent cold work material was precluded. In cases where additional specifications were not clearly identified, a conservative posture was selected to consider the component as being cold worked for the purposes of this assessment. Category 4 components were already assumed to have the potential for cold work in the MRP-191 generic assessments.

The detailed evaluation for NextEra Energy Seabrook cold work assessments concluded that the

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 1/ Page 7 plant-specific material fabrication and design was consistent with the MRP-191 basis, and that the MRP-227-A sampling inspection aging management requirements as related to cold work are directly applicable to NextEra Energy Seabrook.

NextEra Energy Seabrook Evaluation for Question 2 Westinghouse has evaluated the NextEra Energy Seabrook reactor internals components with regard to fuel designs and fuel management according to industry guideline provided in MRP 2013-025.

NextEra Energy Seabrook has not utilized atypical fuel design or fuel management that could make the assumptions of MRP-227-A regarding core loading/core design non-representative for that unit, including power changes/uprates that have occurred over the operating lifetime of the unit. This conclusion is based on comparisons of the NextEra Energy Seabrook core geometry and operating characteristics with the MRP-227-A applicability guidelines for Westinghouse-designed reactors specified in MRP 2013-025.

Specifically, the following comparisons with the MRP-227-A applicability guidelines in MRP 2013-025 were established for the key reactor internals components at NextEra Energy Seabrook.

Components Located Beyond the Outer Radius of the Reactor Core Guideline I - The reactor has been operated with out-in (high-leakage) fuel management for thirty effective full-power years or less and all future operation will use low-leakage fuel management.

Comparison - NextEra Energy Seabrook initiated low-leakage fuel management strategy in Cycle 4 following 2.99 effective full-power years (EFPY) of operation and has been implementing low leakage core designs since that time. There are no current plans to return to out-in (high-leakage) fuel management.

Guideline 2 - For operation going forward, the average power density of the reactor core as defined in MRP 2013-025 shall remain less than 124 W/cm3 .

Comparison - For the last six operating fuel cycles (Cycles 12 through 17), NextEra Energy Seabrook has been operating at a rated power level of 3648 MWt. For the 193 fuel assembly core geometry, the 3648 MWt power level corresponds to a core power density of 111.8 W/cm 3 . This level of power generation is also representative of anticipated future operation.

Guideline 3 - For operation going forward, the nuclear heat generation rate figure of merit (HGR-FOM) as defined in MRP 2013-025 shall not exceed 68 W/cm3 .

Comparison - For the last six operating fuel cycles (Cycles 12 through 17) at NextEra Energy

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 1/Page 8 Seabrook, the range of HGR-FOM at key baffle locations has been less than 68 W/cm 3 and this range is representative of anticipated future operation.

Components Located Above the Reactor Core Guideline 1 - Considering the entire operating lifetime of the reactor, the average power density of the core, as defined in MRP 2013-025 shall remain less than 124 W/cm 3 for a period of more than two effective full-power years.

Comparison - Over the operating lifetime of the NextEra Energy Seabrook reactor, the rated core power level, including power uprates, has varied between 3411 MWt and 3648 MWt. This variation of rated power level corresponds to a power density range of 104.5 W/cm 3 to 111.8 W/cm 3 .

Guideline 2 - Considering the entire operating lifetime of the reactor, the distance between the top of the active fuel stack and the bottom of the upper core plate (UCP) shall not be less than 12.2 inches for a period of more than two effective full-power years.

Comparison - For the NextEra Energy Seabrook reactor internals and fuel assembly geometry, the nominal distance between the top of the active fuel stack and the bottom of the upper core plate (UCP) averaged over the first 17 fuel cycles of operation was not less than 12.2 inches. During that period of time, the nominal distance between the UCP and the top of the active fuel was not less than 12.2 inches for an operating period of more than two effective full-power years.

Components Located Below the Reactor Core Based on the discussion provided in MRP 2013-025, plant-specific applicability of MRP-227-A for components located below the reactor core with no further evaluation required is demonstrated by meeting the MRP-227-A, Section 2.4.

Conclusion NextEra Energy Seabrook complies with Applicant/Licensee Action Item 1 of the NRC SE regarding MRP-227, Revision 0. Therefore, the requirement is met for application of MRP-227-A as a strategy for managing age-related material degradation in the RVI components.

Applicant/Licensee Action Item 2: PWR Vessel Internal Components within the Scope of License Renewal (NRC SE Section 4.2.2)

NRC Applicant/Licensee Action Item 2: PWR Vessel Internal Components within the Scope of License Renewal text in MRP-227-A (Ref. 5) states:

"As discussed in Section 3.2.5.2 of this SE, consistent with the requirementsaddressedin 10 CFR 54.4, each applicant/licenseeis responsiblefor identifying which R VI components are

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 1/Page 9 within the scope ofLR for its facility. Applicants/licenseesshall review the information in Tables 4-1 and 4-2 in MRP-189, Revision 1, and Tables 4-4 and 4-5 in MRP-191 and identify whether these tables contain all of the R VI components that are within the scope of LR for theirfacilities in accordancewith 10 CFR 54.4. If the tables do not identify all the R VI components that are within the scope of LR for its facility, the applicant or licensee shall identify the missing component(s) andpropose any necessary modifications to the program defined in MRP-22 7, as modified by this SE, when submitting its plant-specific NRC-AMP. The NRC-AMP shallprovide assurance that the effects of aging on the missing component(s) will be managedfor the period of extended operation. This issue is Applicant/Licensee Action Item 2. "

NextEra Energy Seabrook Response to Applicant/Licensee Action Item 2 This action item requires comparison of the NextEra Energy Seabrook RVI components that are within the scope of license renewal for NextEra Energy Seabrook to those components contained in MRP-191, Table 4-4, as NextEra Energy Seabrook has a Westinghouse plant design. The components required to be in the NextEra Energy Seabrook aging management program, as described in the PWR Vessel Internals Inspection Plan (Enclosure 2 of this Supplement Letter),

are consistent with those contained in MRP- 191, Table 4.4. No additional components were identified in the NextEra Energy Seabrook design.

Several components have different materials than that specified in MRP- 191, but the differences have no effect on the recommended MRP aging management strategy, or aging is already managed by an alternate NextEra Energy Seabrook aging management program (Ref. Table 2 attached to the end of this enclosure). NextEra Energy Seabrook follows the program strategy in MRP-227-A, with the exception of the flux thimble tubes. As stated in the LRA (Ref. 1, Appendix B, subsection B.2.0) and in responses to RAI 3.1.1.60-01 and 3.1.160-02 (dated February 3,2011), RAI 3.1.60-01/02 (dated April 22, 2011), and RAI 3.1.1.60-02 (dated November 2, 2011), NextEra Energy Seabrook does not credit the flux thimble tube inspection program for aging management. Additionally, the SER to the LRA states that the NRC staff found NextEra Energy Seabrook has provided an acceptable basis that a Flux Thimble Tube Inspection Program is not needed to manage wear or cracking of the NextEra Energy Seabrook flux thimble tubes (Ref. 12, Subsection 3.1.2.1.1). This supports the requirement that the NRC-AMP shall provide assurance that the effects of aging on the NextEra Energy Seabrook RVI components within the scope of license renewal will be managed for the period of extended operation.

The generic scoping and screening of the RVI, as summarized in MRP- 191 and MRP-232, to support the inspection sampling approach for aging management of the RVI specified in MRP-227-A, are applicable to NextEra Energy Seabrook with the exception of the aging management strategy of the NextEra Energy Seabrook flux thimble tubes.

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 1/Page 10 Conclusion NextEra Energy Seabrook complies with Applicant/Licensee Action Item 2 of the NRC SE on MRP-227, Revision 0, and therefore, meets the requirement for application of MRP-227-A as a strategy for managing age-related material degradation in reactor internals components.

Applicant/Licensee Action Item 3: Evaluation of the Adequacy of Plant-Specific Existing Programs (NRC SE Section 4.2.3)

NRC Applicant/Licensee Action Item 3: Evaluation of the Adequacy of Plant-Specific Existing Programs text in MRP-227-A states:

"As addressedin Section 3.2.5.3 in this SE, applicants/licenseesof CE and Westinghouse are requiredto perform plant-specific analysis either to justify the acceptability of an applicant's/licensee'sexisting programs,or to identify changes to the programs that should be implemented to manage the aging of these components for the periodof extended operation. The results of this plant-specific analyses and a descriptionof the plant-specificprograms being relied on to manage aging of these components shall be submitted as part of the applicant's/licensee'sAMP application.The CE and Westinghouse components identifiedfor this type ofplant-specific evaluation include: CE thermal shield positioningpins and CE in-core instrumentationthimble tubes (Section 4.3.2 in MRP-22 7), and Westinghouse guide tube support pins (splitpins) (Section 4.3.3 in MRP-22 7). This is Applicant/Licensee Action Item 3."

NextEra Energy Seabrook Response to Applicant/Licensee Action Item 3 In 2006, the original Alloy X-750 Control Rod Guide Tube (CRGT) split pins were proactively replaced at NextEra Energy Seabrook with Westinghouse designed cold worked 316 stainless steel split pins to mitigate the concern for potential stress corrosion cracking inherent to Alloy X-750 material. As stated in MRP-232 (Material Reliability Program: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internals) Section 4.2.5.2, "The most reliable management approachfor eliminating concerns over guide tube supportpin cracking is a proactive replacementwith Type 316 CW SS supportpins." MRP-232 Section 4.2.5.2 also states that "It should be recognized that crackedguide tube supportpins do not challenge safe plant operation.Even when pins are cracked, the design of the guide tube and the geometry of the pins maintain control rodfunctionality " The main issue with the failure of the CRGT split pins was determined to be potential damage from the loose parts.

Cold-worked Type 316 SS split pins have been installed at other plants since 1997 and none of these plants have experienced any failures. Since other plants have installed split pins since 1997 and Seabrook Station did not install them until 2006, it is reasonably assumed that the other plants will provide a leading indicator. Additionally, there is no specific requirement from the industry working group and/or Westinghouse to inspect the cold worked 316 stainless steel CRGT split pins. Therefore, no inspections of split pins are currently planned. However, as part

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 1/Page 11 of the operating experience review process, the need for inspections will be reevaluated if failures of split pins of the same material occurred in other PWRs.

Applicant/Licensee Action Item 4: B&W Core Support Structure Upper Flange Stress Relief (NRC SE Section 4.2.4)

NRC Applicant/Licensee Action Item 4: B&W Core Support Structure Upper Flange Stress Relief text in MRP-227-A states:

"As discussed in Section 3.2.5.4 of this SE, the B& W applicants/licenseesshall confirm that the core support structure upperflange weld was stress relieved during the originalfabricationof the Reactor Pressure Vessel in order to confirm the applicabilityofMRP-227, as approved by the NRC to theirfacility. If the upperflange weld has not been stress relieved, then this component shall be inspected as a "Primary" inspection category component. If necessary, the examination methods andfrequencyfor non-stress relieved B& W core support structure upper flange welds shall be consistent with the recommendations in MRP-227, as approved by the NRC, for the Westinghouse and CE upper core support barrel welds. The examination coverage for this B& Wflange weld shall conform to the staff's imposed criteriaas described in Sections 3.3.1 and 4.3.1 of this SE. The applicant's/licensee'sresolution of this plant-specific action item shall be submitted to the NRCfor review and approval.This is Applicant/Licensee Action Item

4. "

NextEra Energy Seabrook Response to Applicant/Licensee Action Item 4 This item pertains to B&W Core Support Structure Upper Flange Stress Relief issue. NextEra Energy Seabrook has a Westinghouse design plant. Therefore, this Applicant/Licensee Action Item is not applicable.

Applicant/Licensee Action Item 5: Application of Physical Measurements as part of I&E Guidelines for B&W, CE, and Westinghouse RVI Components (NRC SE Section 4.2.5)

NRC Applicant/Licensee Action Item 5: Application of Physical Measurements as part of I&E Guidelines for B&W, CE, and Westinghouse RVI Components text in MRP-227-A states:

"As addressedin Section 3.3.5 in this SE, applicants/licenseesshall identify plant-specific acceptance criteriato be appliedwhen performing the physical measurements requiredby the NRC-approvedversion of MRP-227for loss of compressibilityfor Westinghouse hold down springs, and for distortion in the gap between the top and bottom core shroudsegments in CE units with core barrelshrouds assembled in two vertical sections. The applicant/licenseeshall include its proposedacceptance criteriaand an explanation of how the proposed acceptance criteriaare consistent with the plants' licensing basis and the need to maintain the functionality of the component being inspected under all licensing basis conditions of operationduring the period of extended operation as part of their submittal to apply the approvedversion ofMRP-

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 1/ Page 12 227. This is Applicant/Licensee Action Item 5."

NextEra Energy Seabrook Response to Applicant/Licensee Action Item 5 NextEra Energy Seabrook is a Westinghouse designed plant and use hold down springs fabricated from 403 stainless steel (Ref. 10). The requirement to perform physical measurements of the hold down spring specified in MRP-227-A, Table 5-3 is only applicable to hold down springs fabricated from 304 stainless steel. Therefore, this item is not applicable to NextEra Energy Seabrook.

Applicant/Licensee Action Item 6: Evaluation of Inaccessible B&W Components (NRC SE Section 4.2.6)

NRC Applicant/Licensee Action Item 6: Evaluation of Inaccessible B&W Components text in MRP-227-A states:

"As addressedin Section 3.3.6 in this SE, MRP-22 7 does not propose to inspect the following inaccessible components. the B& W core barrelcylinders (including vertical and circumferential seam welds), B& Wformer plates, B& W external baffle-to-baffle bolts and their locking devices, B& W core barrel-to-formerbolts and their locking devices, and B& W core barrel assembly internal baffle-to-baffle bolts. The MRP also identified that although the B& W core barrel assembly internalbaffle-to-baffle bolts are accessible, the bolts are non-inspectable using currently available examination techniques.

Applicants/licensees shalljustify the acceptability of these componentsfor continued operation through the period of extended operation by performing an evaluation, or by proposinga scheduled replacement of the components. As part of their applicationto implement the approvedversion of MRP-22 7, applicants/licenseesshallprovide theirjustificationfor the continued operabilityof each of the inaccessible components and, if necessary,provide their planfor the replacement of the componentsfor NRC review and approval. This is Applicant/Licensee Action Item 6."

NextEra Energy Seabrook Response to Applicant/Licensee Action Item 6 This item pertains to B&W Core Support Structure Upper Flange Stress Relief issue. NextEra Energy Seabrook has a Westinghouse design plant. Therefore, this Applicant/Licensee Action Item is not applicable.

Applicant/Licensee Action Item 7: Plant-Specific Evaluation of CASS Materials (NRC SE Section 4.2.7)

NRC Applicant/Licensee Action Item 7: Plant-Specific Evaluation of Cast Austenitic Stainless Steel (CASS) materials text in MRP-227-A states:

"As discussed in Section 3.3.7 of this SE. the applicants/licenseesof B& W, CE, and

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 1/Page 13 Westinghouse reactorsare requiredto develop plant-specific analyses to be appliedfor their facilities to demonstrate that B& W IMI guide tube assembly spiders and CRGT spacer castings, CE lower support columns, and Westinghouse lower support column bodies will maintain their functionality during the periodof extended operation orfor additionalR VI components that may be fabricatedfrom CASS, martensiticstainless steel or precipitationhardenedstainless steel materials. These analyses shall also consider the possible loss offracture toughness in these components due to thermal and irradiationembrittlement, and may also need to consider limitations on accessibilityfor inspection and the resolution/sensitivityof the inspection techniques. The requirement may not apply to components that were previously evaluated as not requiringaging management during development ofMRP-22 7. That is, the requirement would apply to comnponentsfabricatedfromsusceptible materialsfor which an individuallicensee has determined aging management is required,for example during their review performed in accordancewith Applicant/Licensee Action Item 2. The plant-specific analysis shall be consistent with the plant 's licensing basis and the need to maintain the functionality of the components being evaluated under all licensing basis conditions of operation. The applicant/licenseeshall include the plant-specific analysis as part of their submittal to apply the approvedversion ofMRP-22 7. This is Applicant/Licensee Action Item 7. "

NextEra Energy Seabrook Response to Applicant/Licensee Action Item 7 Applicant/Licensee Action Item 7, from the Nuclear Regulatory Commission's (NRC's) final safety evaluation (SE) on MRP-227, Revision 0, notes that, for assessment of CASS materials, the applicant/licensee for license renewal may apply the criteria in the NRC letter of May 19, 2000, License Renewal Issue No. 98-0030, "Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components" (Ref. 13) as the basis for determining whether the CASS materials are susceptible to the thermal aging mechanism. If the application of the applicable screening criteria for the components' material demonstrates that the components are not susceptible to either thermal embrittlement (TE) or irradiation embrittlement (IE), or to the synergistic effects of TE and IE combined, then no other evaluation would be necessary. The NextEra Energy Seabrook CASS reactor vessel internals components and the assessment of their susceptibility to TE are summarized in the table shown below.

The NextEra Energy Seabrook lower internals assembly - bottom-mounted instrumentation (BMI) column cruciforms are ASME SA-35 1, Grade CF8 material. The elemental percentages, from the certified material test reports (CMTRs) for the CASS BMI column cruciform, are input into Hull's formula per the guidance of NUREG/CR-4513 (Ref. 14) to calculate the delta ferrite content of the CASS material. The CMTRs do not list the element percentage for nitrogen; thus, per the guidance of NUREG/CR-4513, nitrogen is assumed to be 0.04%. The CMTRs do not list an elemental percentage for molybdenum. SA-35 1, Grade CF8 did not have a requirement for percent molybdenum in 1971. The 2013 Edition of the American Society of Mechanical Engineers (ASME) Code has SA-35 1, Grade CF8 chemistry requirements that specify a maximum of 0.5 percent molybdenum (Ref. 19); thus, this maximum value is input into Hull's

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 1/ Page 14 formula.

There are a total of 26 BMI column cruciforms, 25 standard cruciforms and 1 special cruciform.

As summarized in the Table shown below, the ferrite content is less than or equal to 20 percent; thus, based on the NRC criteria (Ref. 13), the 26 NextEra Energy Seabrook CASS BMI column cruciforms are not susceptible to TE.

The NextEra Energy Seabrook reactor internals hold-down spring was manufactured from 403 SS, a martensitic stainless steel (Ref. 10).

No martensitic precipitation hardened stainless steel (PH-SS) components were identified for the NextEra Energy Seabrook reactor vessel internals.

Summary of NextEra Energy Seabrook CASS Components and Their Susceptibility to Thermal Embrittlement CASS Components Molybdenum Casting Calculated Susceptibility to TE Content Method Ferrite (Based on Ref. 12)

Content Lower Internals Assembly, 0.5% Maximum Static <20%

• 26 of 26 BMI column cruciforms (Standard and Special) Not Susceptible Note:

1. Ferrite content is based on CMTR chemistry data, nitrogen 0.04 percent, and molybdenum 0.5 percent.

Applicant/Licensee Action Item 8: Submittal of Information for Staff Review and Approval (NRC SE Section 4.2.8)

NRC Applicant/Licensee Action Item 8: Submittal of Information for Staff Review and Approval text in MRP-227-A states:

"As addressedin Section 3.5.1 in this SE, applicants/licenseesshall make a submittalfor NRC review andapproval to credit their implementation of MRP-22 7, as amended by this SE, as an AMP for the R VI components at theirfacility. This submittal shall include the information identified in Section 3.5.1 of this SE. This is Applicant/Licensee Action Item 8."

NextEra Energy Seabrook Response to Applicant/Licensee Action Item 8

1. The revised aging management program that addresses the ten program elements as defined in LR-ISG-2011-04 (Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors) is submitted in Enclosure 2 of this supplement letter. It should be noted that the PWR Vessel Internal Program that was previously submitted in SBK-L-14089, dated June 24, 2014 is superseded by this by this supplement letter.

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 1/Page 15

2. The pressurized water reactor (PWR) vessel internals inspection plan with plant-specific activities for the primary components, expansion components, existing program components, and examination acceptance and expansion criteria are provided in Tables 1 through 4 of the PWR Vessel Internal Program submitted in Enclosure 2 of this supplement letter. The NextEra Energy Seabrook PWR Vessel Internals inspection plan is consistent with the guidance provided in MRP-227-A.

Note 1: The Internals Hold Down Springs (Alignment and Interfacing Components) at NextEra Energy Seabrook are fabricated from 403 stainless steel. The requirement to perform physical measurements of the hold down spring specified in MRP-227-A, Table 5-3 is only applicable to hold down springs made from 304 stainless steel.

Therefore, this item is not applicable to NextEra Energy Seabrook.

Note 2: NextEra Energy Seabrook does not utilize a Flux Thimble Tube Inspection Program for the Flux Thimble Tubes (Bottom Mounted Instrumentation System) because of the double-concentric thimble tube design fabricated from wear resistant, seamless nickel alloy material (INCONEL 600) as discussed in the NRC SER with Open Items for the NextEra Energy Seabrook's LRA, Subsection 3.1.2.1.1 (Ref. 12).

3. The revised UFSAR Supplement for LRA Appendix A, Section A.2.1.7 (PWR Vessel Internals Program) and changes to the commitments related to the PWR Vessel Internals Program are provided in Enclosure 3 of this supplement letter.

4. No technical specification changes are required for NextEra Energy Seabrook based on MRP-227-A and the associated NRC Safety Evaluation.

5. Reactor vessel internals TLAAs are addressed in NextEra Seabrook LRA Section 4.3.3. As discussed in Section 4.3.3, and as amended by response to RAI 4.3.3-1 dated February 3, 2011, and response to RAI 4.3.3-1c dated June 2, 2011, the Seabrook Station Reactor Vessel Internals were designed and constructed prior to the development of ASME Code requirements for core support structures. The reactor coolant system functional design requirements however, were considered in the design. The Reactor Vessel Internals were further analyzed for fatigue as part of the Seabrook Station power uprate and determined that cumulative usage factors would remain less than 1.0. Additionally, as discussed in LRA Section A.2.4.2.2.2, the Metal Fatigue of Reactor Coolant Pressure Boundary Program will monitor the number of design cycles assumed in the fatigue analysis to assure that these will not be exceeded during the period of extended operation per 10 CFR 54.21 (c)(1)(iii). The PWR vessel Internals Program as enhanced by MRP-227-A will manage the aging effects including cracking, loss of material, changes in dimensions, loss of fracture toughness, and loss of preload of the Reactor Vessel Internals components for the period of extended operation per 10 CFR 54.21 (c)(1)(iii).

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 1/ Page 16 Table 2: Material Comparison Table NextEra Energy Seabrook RVI Material Differences From MRP-191 Assembly Subassembly Component Material MRP-191 CF8 Mixing Devices Mixing Devices Upper Internals Seabrook Station 304 SS Assembly Upper Support Column MRP-191 CF8 Column Assemblies Bases Seabrook Station 304 SS MRP-191 304SS Flux Thimble Tubes Plugs Seabrook Station 316SS Flux Thimbles Seabrook Station Alloy 600 (Tubes)

Flux Thimbles MRP-191 316SS (Tubes) Seabrook Station Alloy 600 Lower Internals Lower Support Lower Support MRP-191 304 SS Assembly Column Lower Sort Assemblies Column Bolts Seabrook Station 316 SS Neutron Panel/Thermal Neutron Panel MRP-191 304 SS Lock Caps Shield Seabrook Station 304L SS Radial MRP-191 304SS Radial Support Support Key Keys Bolts Seabrook Station 316 SS

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 1/ Page 17 REFERENCES

1. EPRI Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191), Technical Report 1013234, November, 2006.

2. EPRI Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev. 0), Technical Report 1016596, December, 2008.

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

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 NRC Safety Evaluation Report for EPRI Materials Reliability Program:

Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A),

December 16, 2011.

7. LR-ISG-2011-04, "Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors", May 28, 2013.

8. NextEra Energy Seabrook "Updated Final Safety Analysis Report," Section 4.3, Rev. 10.

9. NextEra Energy Seabrook "Updated Final Safety Analysis Report," Table 5.1-1, Rev. 12 and Table 5.3-1, Rev. 8.

10. NextEra Energy Seabrook "Updated Final Safety Analysis Report," Table 5.2-4, Rev. 8.

11. EPRI Materials Reliability Program: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internals (MRP-232, Rev. 1), Technical Report 1021029, February 2012.

12. NRC Letter, "Safety Evaluation Report with Open Items Related to the License Renewal of Seabrook Station," June 8, 2012 (Accession Number ML12053A192).

13. NRC Letter, "License Renewal Issue No. 98-0030, "Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components"," May 19, 2000 (Accession Number ML003717179).

14. U.S. Nuclear Regulatory Commission, NUREG/CR-4513, Rev. 1, "Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems," August 1994 (Accession Number ML052360554).

15. U.S. NRC Presentation, "Status of MRP-227-A Action Items 1 and 7," June 5, 2013. (NRC ADAMS Accession No. ML13154A152)

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 1/Page 18

16. U.S. NRC Letter, "Summary of February 25, 2013 Telecom with the Electric Power Research Institute and Westinghouse Electric Company," March 15, 2013. (NRC ADAMS Accession No. ML13067A262)

17. EPRI Letter, MRP 2013-025, "MRP-227-A Applicability Template Guideline," October 14, 2013.

18. Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-175). EPRI, Palo Alto, CA: 2005. 1012081.

19. ASME Boiler and Pressure Vessel Code,Section II, 2013 Edition, American Society of Mechanical Engineers.

Enclosure 2 to SBK-L-15073 Revised NextEra Energy Seabrook PWIR Vessel Internals Program Including the PWR Vessel Internals Inspection Plan

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 2/Page 2 13.2.1.7 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-15073 / 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 1 - 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 Alloy X-750 Guide Tube Assembly (GTA) support pins with cold worked (CW) 316 stainless steel support pins during Refueling Outage 11 (Fall of 2006). This modification is considered a preventative action as the CW 316 stainless steel support pins are considered to be much more resistant to the degradation mechanisms of concern for the RVI components than the Alloy X-750 material. Therefore, there

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 2/Page 4 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 of IE is loss of fracture toughness.

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 2/ Page 5 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-15073 / 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 ProgramComponents:

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 NextEra Energy Seabrook.

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 2/Page 7 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. 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 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

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 2/ Page 8 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- 17096, "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, 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

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 2/ Page 9 the RVI components.

Element 7 - Corrective Actions Corrective actions following the detection of unacceptable conditions are fundamentally provided for in the NextEra 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 Aging Management Program".

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 2/ Page 10 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 Progranm is consistent with NUREG-1801 XI.M 16A 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.

PWR Vessel Internals Inspection Plan Table 1 - Westinghouse Plants Existing Programs Components Applicable to NextEra Energy Seabrook Table 2 - Westinghouse Plants Primary Components Applicable to NextEra Energy Seabrook Table 3 - Westinghouse Plants Expansion Components Applicable to NextEra Energy Seabrook Table 4 - Westinghouse Plants Examination, Acceptance, and Expansion Criteria Applicable to NextEra Energy Seabrook

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 2/Page 11 Table 1 - Westinghouse Plants Existing Program Components Applicable to NextEra Energy Seabrook (Based on MRP-227-A, Table 4.9)

(Sheet 1 of 1)

Component Effect Reference Document Generic Requirement Examination Method and (Note 2) (Mechanism) Description Frequency Core Barrel Assembly 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 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 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 Loss of material ASME Code Visual (VT-3) examination All accessible surfaces; one time per Lower core plate (wear) Section XI interval Bottom Mounted Instrumentation System N/A N/A N/A N/A Flux thimble tubes (Note 3)

Alignment and Interfacing Loss of material ASME Code Visual (VT-3) examination All accessible surfaces; one time per Components (wear) Section XI interval Clevis insert bolts (Note 1)

Alignment and Interfacing 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 for the Flux Thimble Tubes (Bottom Mounted Instrumentation System) because of the double-concentric thimble tube design fabricated from wear resistant, seamless nickel alloy material (INCONEL 600) as discussed in the NRC SER with Open Items for the NextEra Energy Seabrook's LRA, Subsection 3.1.2.1.1 (June 8, 2012, Accession Number ML12053A192).

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 2/ Page 12 Table 2 - Westinghouse Plants Primary Components Applicable to NextEra Energy Seabrook (Based on MIRP-227-A, Table 4.3)

(Sheet I of 4)

Component Effect Expansion Link Examination Examination (Mechanism) (Note 1) Method Coverage Control Rod Guide 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 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 colunm bodies, presence of crack-like surface the individual peripheral assemblies (Note (IE and TE) Lower support flaws in flange welds no later 2) column 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 Cracking (SCC) Lower support Periodic enhanced visual 100% of one side of the accessible surfaces Upper core barrel flange column bodies (EVT-1) examination, with of the selected weld and adjacent base weld (non-cast) 10-year intervals, no later than 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 Cracking (SCC, Upper and lower Periodic enhanced visual 100% of one side of the accessible surfaces Upper and lower core IASCC, Fatigue) core barrel cylinder (EVT-1) examination, with of the selected weld and adjacent base barrel cylinder girth axial welds 10-year intervals, no later than 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-15073 / Enclosure 2/Page 13 Table 2 - Westinghouse Plants Primary Components Applicable to NextEra Energy Seabrook (Based on MRP-227-A, Table 4.3)

(Sheet 2 of 4)

Effect Expansion Link Examination Examination Component (Mechanism) (Note 1) Method Coverage Core Barrel Assembly Cracking (SCC, None Periodic enhanced visual 100% of one side of the accessible Lower core barrel flange Fatigue) (EVT-l) 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 Assembly N/A None N/A N/A Baffle-edge bolts (Note 7)

Baffle-Former 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-15073 / Enclosure 2/ Page 14 Table 2 - Westinghouse Plants Primary Components Applicable to NextEra Energy Seabrook (Based on MRP-227-A, Table 4.3)

(Sheet 3 of 4)

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

• Abnormal interaction of distortion, with bolts and indirect effects with fuel assemblies baseline of void swelling in examination former plates) ° Gaps along high between 20 and 40 fluence baffle joint EFPY and (Note 7) subsequent displacement of examinations on a baffle plates near ten-year interval high fluence joint

• Broken or damaged edge bolt locking systems along high fluence baffle joint (Note 8)

Alignment and Interfacing N/A None N/A N/A Components Internals hold down spring (Note 9)

Thermal Shield N/A None N/A N/A Flexures (Note 10)

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 2/ Page 15 Table 2 - Westinghouse Plants Primary Components Applicable to NextEra Energy Seabrook (Based on MRP-227-A, Table 4.3)

(Sheet 4 of 4)

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.

7. Baffle-edge bolts are not applicable to the NextEra Energy Seabrook design.

8. Broken or damaged edge bolt locking systems along high fluence baffle joint is not applicable to the NextEra Energy Seabrook design

9. The Internals Hold Down Springs (Alignment and Interfacing Components) at NextEra Energy Seabrook are fabricated from 403 stainless steel. The requirement to perform physical measurements of the hold down spring specified in MRP-227-A, Table 5-3 is only applicable to hold down springs made from 304 stainless steel. Therefore, this item is not applicable to NextEra Energy Seabrook.

10. NextEra Energy Seabrook utilizes neutron panels design as opposed to the thermal shield flexures.

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 2/ Page 16 Table 3 - Westinghouse Plants Expansion Components Applicable to NextEra Energy Seabrook (Based on MRP-227-A, Table 4.6)

(Sheet I of 2)

Effect Component (Mechanism) Primary Link Examination Method Examination Coverage Upper Internals Assembly Cracking (fatigue, CRGT lower Enhanced visual (EVT-l) 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 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 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 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 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 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-15073 / Enclosure 2/ Page 17 Table 3 - Westinghouse Plants Expansion Components Applicable to NextEra Energy Seabrook (Based on MIRP-227-A, Table 4.6)

(Sheet 2 of 2)

Component Effect Primary Link Examination Method Examination Coverage (Mechanism)

Lower Support Assembly 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 Lower support column bodies N/A N/A N/A N/A (cast)

(Note 3)

Bottom Mounted 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 Bottom-mounted of completely (CRGT) lower difficulty of insertion/withdrawal flux thimble insertion/withdrawal 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).

3. Lower support column bodies are non-cast at NextEra Energy Seabrook.

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 2/ Page 18 Table 4 - Westinghouse Plants Examination Acceptance and Expansion Criteria Applicable to NextEra Energy Seabrook (Based MRP-227-A, Table 5.3)

(Sheet 1 of 6)

Examination Acceptance Link(s) Expansion Criteria Additional Examination Item Criteria (Note 1) Expansion Acceptance Criteria Control Rod Guide Tube 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 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- I 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-15073 / Enclosure 2/Page 19 Table 4 - Westinghouse Plants Examination Acceptance and Expansion Criteria Applicable to NextEra Energy Seabrook (Based MRP-227-A, Table 5.3)

(Sheet 2 of 6)

Examination Acceptance Additional Examination Item Criteria (Note 1) Expansion Link(s) Expansion Criteria Acceptance Criteria Core Barrel Assembly 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-1 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- I 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-15073 / Enclosure 2/ Page 20 Table 4 - Westinghouse Plants Examination, Acceptance, and Expansion Criteria Applicable to NextEra Energy Seabrook (Based MRP-227-A, Table 5.3)

(Sheet 3 of 6)

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

The specific relevant condition is a detectable crack-like surface indication Core Barrel Assembly 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-1 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 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-1 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-15073 / Enclosure 2/ Page 21 Table 4 - Westinghouse Plants Examination, Acceptance, and Expansion Criteria Applicable to NextEra Energy Seabrook (Based MRP-227-A, Table 5.3)

(Sheet 4 of 6)

Examination Acceptance Link(s) Criteria Additional Examination Item Criteria (Note 1) Expansion Expansion Acceptance Criteria Baffle-Former Assembly Baffle-edge bolts N/A N/A N/A N/A (Note 3)

Baffle-Former Assembly 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 column bolts and the The examination b. Barrel-former 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 colunm bolts within the next three fuel cycles

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

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 2/ Page 22 Table 4 - Westinghouse Plants Examination Acceptance and Expansion Criteria Applicable to NextEra Energy Seabrook (Based MRP-227-A, Table 5.3)

(Sheet 5 of 6)

Item Examination Acceptance Link(s) Expansion Criteria Additional Examination Criteria (Note 1) Expansion Acceptance Criteria Baffle-Former Assembly 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 (Note 4)

Alignment and Interfacing N/A None N/A N/A Components Internals hold down spring (Note 5)

United States Nuclear Regulatory Commission SBK-L- 15073 / Enclosure 2/ Page 23 Table 4 - Westinghouse Plants Examination Acceptance and Expansion Criteria Applicable to NextEra Energy Seabrook (Based MRP-227-A, Table 5.3)

(Sheet 6 of 6)

Examination Acceptance Additional Examination Item Criteria (Note 1) Expansion Link(s) Expansion Criteria Acceptance Criteria Thermal Shield Assembly N/A None N/A N/A Thermal shield flexures (Note 6)

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.

3. Baffle-edge bolts are not applicable to the NextEra Energy Seabrook design.

4. Broken or damaged edge bolt locking systems along high fluence baffle joint is not applicable to the NextEra Energy Seabrook design.

5. The Internals Hold Down Springs (Alignment and Interfacing Components) at NextEra Energy Seabrook are fabricated from 403 stainless steel. The requirement to perform physical measurements of the hold down spring specified in MRP-227-A, Table 5-3 is only applicable to hold down springs made from 304 stainless steel. Therefore, this item is not applicable to NextEra Energy Seabrook.

6. NextEra Energy Seabrook utilizes neutron panels design as opposed to the thermal shield flexures.

Enclosure 3 to SBK-L-15073 Changes to the UFSAR Supplement (Section A.2.1.7) and Changes to the Commitments Related to the PWR Vessel Internals Program

United States Nuclear Regulatory Commission SBK-L-15073/ Enclosure 3/ Page 2

1. LRA Appendix A, Section A.2.1.7, PWR Vessel Internals, has been revised as follows.

Note: LRA Section A.2.1.7 was previously revised in SBK-L-14089 dated June 24, 2014. It is merely repeated here for completeness and to assist the reviewer in its evaluation.

A.2.1.7, PWR Vessel Internals 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.

2. Commitment # 1 has been completed and a new commitment #90 has been added as follows:

PROGRAM or UFSAR No. COMMITMENT SCHEDULE TOPIC LOCATION Provide confirmation and acceptability of the PWR Vessel implementation of MRP-227-A by addressing the A.2.1.7 Internals plant-specific Applicant/Licensee Action Items Ao2pl1.7 outlined in section 4.2 of the NRC SER.

Implement the PWR Vessel Internals Program.

The program will be implemented in accordance Prior to te period PWR Vessel with MRP-22 7-A (PressurizedWater Reactor

90. Internals InternalsInspection and Evaluation Guidelines) oferten and NEI 03-08 (Guidelinefor the Management of operation MaterialsIssues).

Enclosure 4 to SBK-L-15073 Revised AMR Items for the Reactor Vessel Internals (Revised LRA Table 3.1.2-3)

United States Nuclear Regulatory Commission SBK-L-15073/ Enclosure 4/ Page 2 Table 3.1.2-3 REACTOR VESSEL INTERNALS Summary of Aging Management Evaluation Intended Aging Effect Requiring Aging NUREG Table Component Type Function Management Management Program 1801 Vol.

2 Item 3.X.1 Item Note Alignment and Interfacing Structural Stainless Reactor Coolant PWR Vessel IV.B2-33 components: S Steel and Neutron Loss of Preload Internals (R-108) 3.1.1-27 A, 1 Internals hold down spring SupportStee Flux Alignment and Interfacing Structural Stainless Reactor Coolant PWR Vessel components: S Steel and Neutron Changes in Dimensions Internals None None A, 1 Internals hold down spring uppo Flux Alignment and Interfacing Structural Stainless Reactor Coolant PWR Vessel components: and Neutron Loss of Material Interals V None None A, Internals hold down spring Support Steel Flux Alignment and Interfacing Coolant PiReactor IrnVessel A, 1 components: Surt Steel and Neutron Cracking Inte(als IV.12-40 3.1.1-37 Upper core plate alignment pins Support Steel Flux Water Chemistry (R112) A, 1 Alignment and Interfacing Structural Stainless Reactor Coolant PWR Vessel IV.B2-34 components: Support Steel and Neutron Loss of Material Interals (R-1 15) 3.1.1-63 A, 1 Upper core plate alignment pins Flux Baffle-to- Former Assembly: Structural Reactor Coolant PWR Vessel A, 1 Stainless and Neutron Cracking Internals IV.B2-10 3.1.1-30 Baffle-to-former bolts Support Steel Flux Wackr Water Chemistry (R-125) A,1I1 A,

Baffle-to- Former Assembly: Structural Stainless Reactor Coolant PWR Vessel IV.B2-6 Baffle-to-former bolts Support Steel and Neutron Loss of Fracture Toughness Interals (R-128) 3.1.1-22 A, 1 Flux Baffle-to- Former Assembly: Structural Stainless Reactor Coolant PWR Vessel IV.B2-4 Baffle-to-former bolts Support Steel and Neutron Change in Dimensions Internals (R-126) 3.1.1-33 A, 1 Flux Baffle-to- Former Assembly: Structural Stainless Reactor Coolant PWR Vessel IV.B2-5 Baffle-to-former bolts Support Steel Flux Internals (R-129) 3.1.1-27 A, I

United States Nuclear Regulatory Commission SBK-L-15073/ Enclosure 4/ Page 3 Table 3.1.2-3 REACTOR VESSEL INTERNALS Summary of Aging Management Evaluation Intended Aging Effect Requiring Aging NUREG Table Component Type Iuntene Material Environment A n agEeteqng Management 1801 Vol. 3.X.1 Note Function Management Program 2 Item Item Direct Flow Reactor Coolant Baffle-to- Former Assembly: Stainless and Neutron Change in Dimensions PWR Vessel IV.B2-1 3.1.1-33 A, 1 Baffle and former plates Stctual Steel Flux Internals (R-124)

Support Direct PWR Vessel Flow Stainless Reactor Coolant Interals IV.B2-2 A, I Baffle-to- Former Assembly:

Baffle and former plates Sttul Steel Flux Cracking (R-123) 3.1.1-30 A, 1 Support Water Chemistry Baffle-to- Former Assembly: Structural Stainless Reactor Coolant and Neutron Cracking PWR Vessel A, I 3.1.1-30 Barrel-to-former bolts Support Steel Flux Water Chemistry (R-125) A, 1 Baffle-to- Former Assembly: Structural Stainless Reactor Coolant PWR Vessel IV.B2-6 Barrel-to-former bolts Support Steel Flux Internals (R-128) 3.1.1-22 A, I Baffle-to- Former Assembly: Structural Stainless Reactor Coolant PWR Vessel IV.B2-4 Barrel-to- former bolts Support Steel and Neutron Change in Dimensions Internals (R-126) 3.1.1-33 A, I Flux Baffle-to- Former Assembly: Structural Stainless Reactor Coolant and Neutron Loss of Preload PWR Vessel Internals IV.B2-5 (R- 129) 3.1.1-27 A, 1 Barrel-to-former bolts Support Steel Flux Bottom-mounted instrumentation system: Structural Stainless Reactor Coolant PWR Vessel Bottom-mounted Support Steel and Neutron Cracking Internals None None A, I instrumentation (BMI) column Flux bodies Bottom-mounted instrumentation system: Structural Stainless Reactor Coolant PWR Vessel IV.B2-22 A, I Bottom-mounted Support Steel and Neutron Loss of Fracture Toughness InterVals (R-B141) 3.1.1-22 instrumentation (BMI) column Flux bodies I I I I I I

United States Nuclear Regulatory Commission SBK-L-15073/ Enclosure 4/ Page 4 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 Control rod guide tube (CRGT) Structural Stainless Reactor Coolant PWR Vessel assemblies: and Neutron Loss of Material None None A, 1 CRGT guide plates (cards) Support Steel Flux Internals PWR Vessel A Control rod guide tube (CRGT) Reactor Coolant Internals IV.B2-28 A, assemblies: Surt Steel and Neutron Cracking (R-2118) 3.1.1-37 CRGT lower flange welds Support Steel Flux Water Chemistry A, 1 Control rod guide tube (CRGT) Stainless Reactor Coolant assemblies: and Neutron Loss of Fracture Toughness PWR Vessel IV.B2-22 3.1.1-22 A, 1 CRGT lower flange welds Support (including Flux Internals (R-141)

CASS)

PWR Vessel 1 Control rod guide tube (CRGT) SorA, assemblies: Structural Stainless Reactor Coolant Internals IV.B2-28 31-7 Guide tube support pins (split Support Steel; and Neutron Cracking (R-1 18) 3.1.1-37 pins) Nickel alloy Flux Water ChemistryA, Control rod guide tube (CRGT) Stainless Reactor Coolant assemblies: Structural Stanles Steel; eacorCooan and Neutron Loss of Material PWR Vessel None None Guide tube support pins (split A, 1 Support Nickel alloy Flux Internals pins)

Core barrel assembly: Direct PWR Vessel A, 1 Upper core barrel and lower Flow Stainless Reactor Coolant Cracking 3.1.

core barrel circumferential Steel lux (R- 120) A,1I (grt) elsStructural Flux A, I (girth) welds Water Chemistry Core barrel assembly: Direct Upper core barrel and lower Flow Stainless and Neutron Loss of Fracture Toughness PWR Vessel IV.B2-9 core barrel circumferential Sttul Steel Flux Internals (R-122) 3.1.1-22 A, I (girth) welds Sura Support

United States Nuclear Regulatory Commission SBK-L-15073/ Enclosure 4/ Page 5 Table 3.1.2-3 REACTOR VESSEL INTERNALS Summary of Aging Management Evaluation M aterial Environment AgingMEffect Requiring anagement Aging en Managram NUREG Vol.

180 Vol. Table 3 X1 Not Component Type Function Intended IneddManagement 1801 3.X.1 Note MaagmetProgram 2 Item Item Direct PWR Vessel Core barrel assembly: Flow Stainless Reactor Coolant Internals IV.B2-8 A, 1 Upper core barrel and lower Stael and Neutron Cracking (R-120) 3.1.1-30 core barrel vertical (axial) welds Structural Flux Water Chemistry A, 1 Support Direct Core barrel assembly: Flow Stainless Reactor Coolant PWR Vessel IV.B2-9 Upper core barrel and lower and Neutron Loss of Fracture Toughness Internals (R- 122) 3.1.1-22 A, 1 core barrel vertical (axial) welds Structural Flux Support Direct Core barrel assembly: Flow Stainless Reactor Loss of Material CoolantVessel and Neutron Pone Vessel Core barrel flange St l Steel Flux Internals None None A, I Support Direct PWR Vessel Core barrel assembly: Stainless Reactor Coolant Internals IV.B2-8 A, I Core barrel outlet nozzle welds Stel and Neutron Cracking nel120) 3.1.1-30 Structural Steel Flux Water Chemistry A, 1 Support Direct Core barrel assembly: Flow Stainless Reactor Coolant PWR Vessel IV.B2-9 Core barrel outlet nozzle welds and Neutron Loss of Fracture Toughness Internals (R-122) 3.1.1-22 A, 1 Structural Steel Flux Support Direct PWR Vessel Flow Reactor Coolant A, 1 Core barrel assembly: Stainless Internals IV.B2-8 and Neutron Cracking 3.1.1-30 Lower core barrel flange weld Steel (R-120)

Structural Flux A, 1 Water Chemistry Sunnort

United States Nuclear Regulatory Commission SBK-L-15073/ Enclosure 4/ Page 6 Table 3.1.2-3 REACTOR VESSEL INTERNALS Summary of Aging Management Evaluation Aging Effect Requiring Aging NUREG Table Intended tntene MaManagement Management Program 1801 Vol.

2 Item 3.X.1 Item Note Direct FlwReactor Coolant PWYR Vessel A, 1 Core barrel assembly: Flow Stainless and Neutron Cracking Internals IV.B2-8 31-Upper core barrel flange weld Structural Steel Flux Water Chemistry (R-120) A, I Support Lower internals assembly: Structural Nickel Reactor CoolantVessel and Neutron Loss of Material PWtessel None None A, 1 Clevis insert bolts or screws Support Alloy Flux Internals Lower internals assembly: Structural Nickel Reactor Coolant PWR Vessel IV.B2-14 Clevis insert bolts or screws Support Alloy Flux Internals (R-137)

ReactorCoolantPWrR Vessel A Lower internals assembly: Structural Nickel and Neutron Reactor Coolant Cracking WInternals IV.2-7 IV.2-16 3.1.1-3 A, 1 Clevis insert bolts or screws Support Alloy Flux Water Chemistry (R-133) A, 1 Direct Flowc PWkR Vessel A Lower internals assembly: Flow Stainless Reactor Coolant Internals IV.2-20A, and Neutron Cracking (R-20 3.1.1-30 Lower core plate Structural Steel Flux Water Chemistry A, 1 Support WatrCemitr Direct Lower interals assembly: Flow Stainless Reactor Coolant PWR Vessel IV.B2-18 Lower core plate Steel and Neutron Loss of Fracture Toughness Internals (R-132) 3.1.1-22 A, 1 Structural Flux Support Lower support assembly: Sttul Stainless Reactor Coolant PWR Vessel A, 1 Lower support forging or Support Steel and Neutron Cracking Internals V.132-24 3.1.1-30 casting Flux Water Chemistry (R138) A, 1 Lower support assembly: Structural Stainless Reactor Coolant PWR Vessel IV.B2-17 Lower support forging or S rt Steel and Neutron Loss of Fracture Toughness Internals (R- 135) 3.1.1-22 A, 1 casting Support Steel Flux

United States Nuclear Regulatory Commission SBK-L-15073/ Enclosure 4/ Page 7 Table 3.1.2-3 REACTOR VESSEL INTERNALS Summary of Aging Management Evaluation Intended Aging Effect Requiring Aging NUREG Table Component Type Fnten Material Environment Management 1801 Vol. 3.X.1 Note Function Management Program 2 Item Item Lower support assembly: SWRuV Coolant It el lReactor A, I Lower support column bodies Sura Steel and Neutron Cracking Internals IV.-1324 3.1.1-30 (non-cast) Support Steel Flux Water Chemistry (R-138) A, I Lower support assembly: Structural Stainless Reactor Coolant PWR Vessel IV.B2-9 Lower support column bodies Support Steel and Neutron Loss of Fracture Toughness Internals (R- 122) 3.1.1-22 A, 1 (non-cast) Flux Lower support assembly: Structural Stainless Reactor Coolant PWR Vessel A, 1 Lower support columnbol Stupprt Stnee and Neutron Cracking Internals IV.B2-16 3.1.1-30 Lower support column bolts Support Steel Flux Water Chemistry (R133) A, 1 Lower support assembly: Structural Stainless Reactor Coolant PWR Vessel IV.B2-17 Lower support column bolts Support Steel and Neutron Loss of Fracture Toughness Internals (R-135) 3.1.1-22 A, I Flux Lower support assembly: Structural Stainless Reactor Coolant PWR Vessel IV.B2-25 Lower support column bolts Support Steel and Neutron Loss of Preload Internals (R-136) 3.1.1-27 A, 1

__________Flux Direct Stainless Reactor vessel internal Flow Steel; Reactor Coolant IV.B32-31 components internalal Nickel and Neutron Cumulative fatigue damage TLAA (R-53) 3.1.1-5 A, 1 sStructural N eloy Flux Support Alloy Direct Stainless Reactor vessel internal Flow Steel; Reactor Coolant IV.,2-32 Nickel and Neutron Loss of Material Water Chemistry (RP-24) 3.1.1-83 A, I components sStructural Alloy Flux Support Alloy Reactor vessel internals:

ASMIE Section XI, Examination Direct Reactor Coolant Category B-N-3 core support Flow Steel; Reactor Nuon C Inservice Inspection, IV.Co2-26 N structure components (not Fluxel lnsectioe In c, (R-26 already identified as "Existing Structural Nickel Flux Subsections IWB, (R-142)

Programs" components in Support Alloy IWC, and IWD MRP-227-A)

United States Nuclear Regulatory Commission SBK-L-15073/ Enclosure 4/ Page 8 Table 3.1.2-3 REACTOR VESSEL INTERNALS Summary of Aging Management Evaluation Aging Effect Requiring Aging NUREG Table Intended Component Type unctiond Material Environment Management 1801 Vol. 3.X.1 Note Function Management Program 2 Item Item Reactor vessel internals:

ASMIE Section XI, Examination DirectStil Stainlesss ASME ME Section ec onXXl Category CterucBu B-N-3 Nco core support rensupport Flow Steel; Reactor Coolant Inservice Inspection, IV.B2-26 structure components (not Nickel and Neutron Loss of Material Subsections IWB, (R-142) None A, I already identified as "Existing Structural Alloy Flux IWC, and IWD Programs" components in Support MRP-227-A)

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

United States Nuclear Regulatory Commission SBK-L-15073/ Enclosure 4/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 5 to SBK-L-15073 LRA Appendix A - Updated Final Safety Analysis Report Supplement Table A.3, License Renewal Commitment List Updated to Reflect Changes to Date

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 5/ Page 2 A.3 LICENSE RENEWAL COMMITMENT LIST UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Provide confirmation and acceptability of the implementation of MRP-227- Ma ,, 2015

1. PWR Vessel Internals A by addressing the plant-specific Applicant/Licensee Action Items A.2.1.7 outlined in section 4.2 of the NRC SER. Complete Closed-Cycle Cooling Enhance the program to include visual inspection for cracking, loss of Prior to the period of extended

2. Water material and fouling when the in-scope systems are opened for A.2.1.12 operation.

maintenance.

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 HeavyLoadRltd adLgtoA2l1 oeain Refuelandling structural components and the effects of wear on the rails in the rail system. 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 operation.

Refueling) Handling Systems Compressed Air Enhance the program to include an annual air quality test requirement for A.2.1.14 Prior to the period of extended

5. Monitoring the Diesel Generator compressed air sub system. 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.

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 Prior to the period of extended

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

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 5/ Page 3 Enhance the program to include NFPA 25 (2011 Edition) guidance for Fire Water System "where sprinklers have been in place for 50 years, they shall be replaced or A2.1.16 Prior to the period of extended 9

representative samples from one or more sample areas shall be submitted to operation.

a 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 commence during the ten year period prior to the period of extended operation and continue through the period of extended operation.

Enhance the program to include 1) In-scope outdoor tanks, except fire water storage tanks, constructed on soil or concrete, 2) Indoor large volume storage tanks (greater than 100,000 gallons) designed to near-atmospheric internal pressures, sit on concrete or soil, and exposed internally to water, Aboveground Steel 3) Visual, surface, and volumetric examinations of the outside and inside Within 10 years prior to the period

12. Tanks surfaces for managing the aging effects of loss of material and cracking, 4) A.2.1.17 of extended operation.

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

Enhance the program to perform exterior inspection of the fire water storage tanks annually for signs Z11 of degradation and include an ultrasonic Within ten years prior to the period

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

Protection Water Storage Tanks per the guidance provided in NFPA 25 (2011 Edition).

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 5/ Page 4 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 Prior to the period of extended

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

and to 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

17. Fuel Oil Chemistry tank bottom during the 10-year draining, cleaning and inspection of the A.2.1.18 Prior to the period of extended Diesel Generator fuel oil storage tanks, Diesel Generator fuel oil day tanks, operation.

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

Reactor Vessel Enhance the program to specify that all pulled and tested 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 Reactor Vessel limitations or bounds defined by the Reactor Vessel Surveillance Program, Prior to the period of extended

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

impact of plant operation changes on the extent of Reactor Vessel embrittlement will be evaluated and the NRC will be notified.

Enhance the program as necessary to ensure the appropriate withdrawal schedule for capsules remaining in the vessel such that one capsule will be 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.

E185-82 and that bounds the 60-year fluence, and the remaining capsule(s) will be removed from the vessel unless determined to provide meaningful metallurgical data.

Enhance the program to ensure that any capsule removed, without the intent Reactor Vessel to test it, is stored in a manner which maintains it in a condition which A.2.1.19 Prior to the period of extended

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

operation.

United States Nuclear Regulatory Commission SBK-L-l15073 / Enclosure 5/ Page 5 Implement the One Time Inspection Program. A.2.1.20 Within ten years prior to the period

22. 22. One-Time Inspection 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 A.2.1.22 the period of extended operation One-Tim eInsp o of Implement the One-Time Inspection of ASME Code Class 1 Small Bore- Within ten years prior to the period

2. ASME Code Class 1 Pip2g.Prgram 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 Prior to the period of extended

26. Monitoring outage inspection frequency, the training requirements for inspectors, and A.2.1.24 operation.

the required periodic reviews to determine program effectiveness.

Inspection of Internal Surfaces in Implement the Inspection of Internal Surfaces in Miscellaneous Piping and A.2.1.25 Prior to the period of extended

27. Miscellaneous Piping and Ducting Components Program. operation.

Ducting Components

28. Lubricating Oil Analysis Enhance the program to add required equipment, lube oil analysis required, A.2.1.26 Prior to the period of extended 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 A.2.1.26 operation.

oil collection tanks prior to the period of extended operation.

ASME Section XI, . A 1 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.

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 5/ Page 6 Structures Monitoring Enhance procedure to include inspection of opportunity when planning A.2.1.31 Prior to the period of extended 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

35. Environmental 50.49 Environmental Qualification Requirements Used in Instrumentation A.2.1.33 prratote Qualification Circuits program.

Requirements Used in Instrumentation Circuits Inaccessible Power Cables Not Subject to 10 CFR 50.49 Implement the Inaccessible Power Cables Not Subject to 10 CFR 50.49 A.2.1.34 Prior to the period of extended

36. Environmental Environmental Qualification Requirements program. operation.

Qualification Requirements Prior to the period of extended A.2.1.35 operaton.

37. Metal Enclosed Bus Implement the Metal Enclosed Bus program. operation.

Prior to the period of extended

38. Fuse Holders I Implement the Fuse Holders Iperation.

program. A.2.1.36 operation.

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 5/ 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 A.2.2.1 Prior to the period of extended

40. 345 KV SF6 Bus Implement the 345 KV SF6 Bus program. operation.

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

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

Boundary Metal Fatigue of Reactor Enhance the program to implement a software program, to count transients A.2.3.1 Prior to the period of extended

42. Coolant PressureIA.31 operation.

Boundary to monitor cumulative usage on selected components.

Pressure -Temperature The updated analyses will be Limits, including Low submitted at the appropriate time to

43. Temperature Seabrook Station will submit updates to the P-T curves and LTOP limits to A.2.4.1.4 comply with 10 CFR 50 Appendix Overpressure Protection the NRC at the appropriate time to comply with 10 CFR 50 Appendix G. G, Fracture Toughness Limits Requirements.

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 5/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 Environmentally- reactor water environment. This includes applying the appropriate Fen At least two years prior to entering

44. Assisted Fatigue factors to valid CUFs determined from an existing fatigue analysis valid for A.2.4.2.3 the period of extended operation.

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

NRC-approved code case).

(2) If acceptable CUFs cannot be demonstrated for all the selected locations, then additional plant-specific locations will be evaluated. For the additional plant-specific locations, 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-15073 / Enclosure 5/ Page 9

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

46. Monitoring and Coordinator and an Inspection Results Evaluator. A.2.1.38 operation.

Maintenance Enhance the program by including, "Instruments and Equipment needed for Protective Coating inspection may include, but not be limited to, flashlight, spotlights, marker A.2.1.38 Prior to the period of extended

47. Monitoring and 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 Prior to the period of extended

49. Monitoring and by the responsible evaluation personnel, who is to prepare a summary of A.2.1.38 operation.

Maintenance findings and recommendations for future surveillance or repair.

Baseline inspections were ASME Section XI, Perform UT of the accessible areas of the containment liner plate in the completed during OR 16. Repeat A.2.1.27 examinations at intervals of no

50. Subsection XI, vicinity of the moisture barrier for loss of material. Perform opportunistic UT of inaccessible areas.

more than five (5) refueling outages.

51. Number Not Used Implement measures to maintain the exterior surface of the Containment A.2.1.28 Ongoing

52. ASME Subsection IWLXI, Section 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 Prior to the period of extended

53. Studs that has a yield strength > 150 ksi with ones that do not exceed 150 ksi. A.2.1.3 operation.

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 5/ 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 Steam Generator Tube perform routine tube-to-tubesheet weld inspections for the remaining life of

54. A.2.1.10 Complete Integrity the steam generators, or

2) Perform an analytical evaluation showing that the structural integrity of the steam generator tube-to-tubesheet interface is adequately maintaining the pressure boundary in the presence 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 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

59. Penetrations applicable Bulletins, Generic Letters, and staff accepted industry A.2.2.3 operation.

guidelines.

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 5/ Page 11 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.

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.2.14 the period of extended operation.

Enhance the program to include a statement that sampling frequencies are A.2.1.2 Prior to the period of extended

62. Water Chemistry increased when chemistry action levels are exceeded. 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

63. Flow Induced Erosion increase in the CVCS Charging Pump mini flow above the acceptance N/A prratote criteria may be indicative of erosion of the mini flow orifice as described in operation.

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-

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

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

65. Flux Thimble Tube Implement measures to ensure that the movable incore detectors are not Prior to entering the period of 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 borated water to be examined for concrete degradation and also

67. Srutres expose rebar to detect any degradation such as loss of material. The A.2.1.31 No later than December 31, 2015.

.Program removed core will also be subjected to petrographic examination for concrete degradation 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 Replace the Diesel Generator Heat Exchanger Plastisol PVC lined Service A.2. 1.11 Complete.

69. Water SystemCooling Open-Cycle Water piping with piping fabricated from AL6XN material.

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 5/ Page 12 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 A.2.1.12 Within ten years prior to the period

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

cooling water system.

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 A.2.1.31A extended operation.

Program acceptance 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

73. Surfaces in provide a representative sample of 20%, or a maximum of 25, of each A.2.1.25 Prior to entering the period of Miscellaneous Piping and identified material, environment, and aging effect combinations during each A 5tentering Ducting 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 Prior to the period of extended

74. Fire Water System damage and installed in the proper orientation (e.g., upright, pendant, or A.2.1.16 sidewall). Any sprinkler that is painted, corroded, damaged, loaded, or in the improper orientation, and any glass bulb sprinkler where the bulb has emptied, will be evaluated for replacement.

Enhance the program to a) conduct an inspection of piping and branch line conditions every 5 years by opening a flushing connection at the end of one main and by removing a sprinkler toward the end of one branch line for the purpose of inspecting for the presence of foreign organic and inorganic material per the guidance provided in NFPA 25 (2011 Edition) and b) If the presence of sufficient foreign organic or inorganic material to obstruct pipe A.2.1.16 Prior to the period of extended

75. Fire Water System or sprinklers is detected during pipe inspections, the material will be operation.

removed and its source is determined and corrected.

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

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 5/ Page 13 Enhance the Program to conduct the following activities annually per the guidance provided in NFPA 25 (2011 Edition). Prior to the period of extended

76. Fire Water System nain m drain tests A.2.1.16operation.

• deluge valve trip tests

• fire water storage tank exterior surface inspections 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. Fire Water System 0 The requirement that if there is a 10 percent reduction in full flow A.2.1.16 Prior to the period of extended pressure when compared to the original acceptance tests or previously operation.

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

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

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

78. Monitoring components exposed to condensation (due to the in-scope component being A.2.1.24 operation.

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

Enhance the program to include visual inspection of internal

79. Open-Cycle Water SystemCooling coatings/linings for loss of coating integrity. A.2. 1.11 Within of 10 years extended prior to the period operation.

Enhance the program to include visual inspection of internal

80. Fire Water System coatings/linings for loss of coating integrity. A.2.1.16 Withn1ea priori of extended operation.

Enhance the program to include visual inspection of internal 81 ulOlCeity coatings/linings for loss of coating integrity. A.2. 1.18Wihn1yerpiotoheeid

81. uel il hemitryof extended operation.

Inspection of Internal Enhance the program to include visual inspection of internal

82. Surfaces in coatings/linings for loss of coating integrity. A.2.1.25 Within 10 years prior to the period Miscellaneous Piping and of extended operation.

Ducting Components

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 5/ 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 Alkali-Silica Reaction Experience Element of the Alkali-Silica Reaction Monitoring Program to Prior to the period of extended

83. Monitoring determine whether there is a need to enhance the program to monitor A.2.1.3 1A operation.

expansion in the out-of-plane 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.

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

Enhance the program to perform additional tests and inspections on the Fire Water Storage Tanks as specified in Section 9.2.7 of NFPA 25 (2011 Prior to the period of extended

85. Fire Water System Edition) in the event that it is required by Section 9.2.6.4, which states A.2.1.16 operation.

"Steel tanks exhibiting signs of interior pitting, corrosion, or failure of coating shall be tested in accordance with 9.2.7."

Enhance the program to include disassembly, inspection, and cleaning of A.2.1.16 Prior to the period of extended

86. Fire Water System the mainline strainers every 5 years. operation.

Increase the frequency of the Open Head Spray Nozzle Air Flow Test from Prior to the period of extended

87. Fire Water System every 3 years to every refueling outage to be consistent with LR-ISG-2012- A.2.1.16 operation.

02, AMP XI.M27, Table 4a.

Enhance the program to include verification that a) the drain holes associated with the transformer deluge system are draining to ensure complete drainage of the system after each test, b) the deluge system drains

88. Fire Water System and associated piping are configured to completely drain the piping, and A.2.1.16 of extended operation.

c) normally-dry piping that could have been wetted by inadvertent system actuations or those that occur after a fire are restored to a dry state as part of the suppression system restoration.

Inspection oflInternal Incorporate Coating Service Level III requirements into the RCP Motor Surfaces in Refurbishment Specification for the internal painting of the motor upper Prior to the period of extended

89. sce s in bearing coolers and motor air coolers. All four RCPs will be refurbished A.2.1.25 operation.

Miscellaneous Piping and and replaced using the Coating Service Level III requirements prior to Ducting Components entering the period of extended operation.

United States Nuclear Regulatory Commission SBK-L-15073 / Enclosure 5/Page 15 Implement the PWR Vessel InternalsProgram. The program will be implemented in accordancewith MRP-22 7-A (PressurizedWater Reactor A.2.1.7

90. PWR Vessel Internals InternalsInspection and Evaluation Guidelines) and NEI 03-08 operation (Guidelinefor the Management of MaterialsIssues).