SBK-L-11123, Response to Request for Additional Information, NextEra Energy Seabrook License Renewal Application, Request for Additional Information - Set 14

Response to Request for Additional Information, NextEra Energy Seabrook License Renewal Application, Request for Additional Information - Set 14

ML11154A133
Person / Time
Site:	Seabrook
Issue date:	06/02/2011
From:	Freeman P NextEra Energy Seabrook
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SBK-L-11123
Download: ML11154A133 (28)

Text

NExTeraM ENERGY ,,ABROK June 2, 2011 SBK-L-1 1123 Docket No. 50-443 U.S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852 Seabrook Station Response to Request for Additional Information NextEra Energy Seabrook License Renewal Application Request for Additional Information - Set 14

References:

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

2. NRC Letter "Request for Additional Information Related to the Review of the Seabrook Station License Renewal Application (TAC NO. ME4028) - Request for Additional Information Set 14," May 23, 2011. (Accession Number MLI 1132A007)

3. NextEra Energy Seabrook, LLC letter SBK-L- 11069, "Seabrook Station, Response to Request for Additional Information, NextEra Energy Seabrook License Renewal Application, Request for Additional Information - Set 12, April 22, 2011.

(Accession Number ML11115A116)

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

74v-4 NextEra Energy Seabrook, LLC, P.O. Box 300, Lafayette Road, Seabrook, NH 03874

United States Nuclear Regulatory Commission SBK-L- 11123 / Page 2 In Reference 2, the NRC requested additional information in order to complete its review of the License Renewal Application. The Enclosure contains NextEra's response to the request for additional information with the exception of RAI B. 1.4-1. Asidiscussed with the Seabrook Station Project Manager, NextEra is participating in industry initiatives with NEI and the NRC related to this issue. The response to RAI B.1.4-1 will be provided by June 27, 2011.

There are no new or revised regulatory commitments contained in this letter. Enclosure 2 provides the current LRA Appendix A - Final Safety Report Supplement Table A.3, License Renewal Commitment List, updated to reflect the license renewal commitment changes made in NextEra's correspondence to date.

If there are any questions or additional information is needed, please contact Mr. Richard R.Cliche, License Renewal Project Manager, at (603) 773-7003.

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

Sincerely, NextEra Energy Seabrook, LLC.

Paul 0. Freeman Site Vice President Enclosure cc:

W.M. Dean, NRC Region I Administrator G. E. Miller, NRC Project Manager, Project Directorate 1-2 W. J. Raymond, NRC Resident Inspector R. A. Plasse Jr., NRC Project Manager, License Renewal M. Wentzel, NRC Project Manager, License Renewal Mr. Christopher M. Pope Director Homeland Security and Emergency Management New Hampshire Department of Safety Division of Homeland Security and Emergency Management Bureau of Emergency Management 33 Hazen Drive Concord, NH 03305 John Giarrusso, Jr., Nuclear Preparedness Manager The Commonwealth of Massachusetts Emergency Management Agency 400 Worcester Road Framingham, MA 01702-5399

United States Nuclear Regulatory Commission SBK-L-1 1123 / Page 3 NEXTera ENERG-7Yl I, Paul 0. Freeman, Site Vice President of NextEra Energy Seabrook, LLC hereby affirm that the information and statements contained within are based on facts and circumstances which are true and accurate to the best of my knowledge and belief.

Sworn and Subscribed Before me this

." day of June, 2011 Paul 0. Freeman Site Vice President Notary Public

United States Nuclear Regulatory Commission SBK-L- 11123 / Enclosure 1 / Page 1 Enclosure 1 Request for Additional Information (RAI) B.2.1.12-9

Background:

By letter dated January 21,2011, the staff issued RAI B.2.1.12-7 requesting that the applicant justify why the Closed-Cycle Cooling Water System Program, which is based on EPRI 1007820, "Closed Cooling Water Chemistry Guideline," does not need to manage microbiologically influenced corrosion (MIC) in the closed cycle cooling water systems. In its response dated February 18, 2011, the applicant stated that the GALL Report does not include any line items for PWRs that include the closed-cycle cooling water environment with MIC as an aging effect, and therefore, it did not consider MIC to be an aging effect requiring management. The applicant also stated that its review of plant-specific operating experience did not identify any MIC issues in the close-cycle cooling water systems, and reiterated that the Closed Cycle Cooling Water System Program does not manage loss of material due to MIC.

The staff noted that the applicant's closed cycle cooling systems for its diesel generator jacket water, fire pump diesel coolant, and control building air handling systems use glycol as a chemical treatment. The staff also noted that MIC is a stated concern in EPRI 1007820 for closed cycle cooling systems utilizing glycol formulations.

Issue:

The applicant did not provide a technical basis for"why loss of material due to MIC does not need to be included as part of the Closed-Cycle Cooling Water System Program. The staffs position-and that stated in EPRI 1007820-is that MIC can occur in closed cycle cooling water systems. The staff further noted that the applicant's lack of plant-specific operating experience associated with MIC may be attributable to the existing additives that mitigate this mechanism. However, as noted in SRP-LR Section A.1.2.1, "Applicable Aging Effects," an aging effect should be identified as applicable for license renewal even if there is a prevention or mitigation program associated with that aging effect.

Request:

Please provide plant-specific data to demonstrate that the lack of problems with MIC at the site cannot be attributed to the existing chemical treatment in the closed cooling water systems or revise the Closed-Cycle Cooling Water System Program to include monitoring for MIC.

NextEra Energy Seabrook Response:

The Seabrook Station Aging Management Program for Closed-Cycle Cooling Water Systems complies with EPRI Technical Report 1007820, "Closed Cooling Water Chemistry Guideline" and takes no exception relative to testing for microbiological activity. As previously stated in response to RAI B.2.1.12-7, the Seabrook Station closed-cycle cooling chemistry control procedure includes testing for biological activity as a diagnostic parameter in keeping with the guidelines provided by

United States Nuclear Regulatory Commission SBK-L- 11123 / Enclosure 1 / Page 2 EPRI. Therefore, Seabrook Station will clarify the intent of the Aging Management Program for Closed-Cycle Cooling Water Systems to specifically state that testing for microbiological activity is a function of this program.

Based on the above discussion, the following changes have been made to the LRA:

1. In Section A.2.1.12, on page A-10, the Ist paragraph is revised to read as follows:

"The Closed-Cycle Cooling Water Program manages aging effects of cracking, loss of material, and reduction of heat transfer in closed cycle cooling water systems. Closed-Cycle Cooling Water (CCCW) systems are described as systems not subject to significant sources of contamination, in which water chemistry is controlled and in which heat is not directly rejected to the ultimate heat sink. The program scope includes activities to manage aging in the Primary Component Cooling Water system and Emergency Diesel Generator Jacket Water cooling systems. The program also includes fire pump diesel engine glycol coolant system, the Control Building Air Handling glycol coolant system (safety-related), and the Thermal Barrier Cooling Water system. Glycol containingsystems within the scope of this programare monitoredfor the presence of microbiologicalactivity in accordancewith the EPRI Closed-Cycle Cooling Water guidelines."

2. In Section B.2.1.12, on page B-78, the 1St full paragraph (following exception "c") is revised to read as follows:

"The fire pump diesel engine glycol coolant system, the Control Building Air Handling glycol coolant system, and the Diesel Generator Jacket Water glycol coolant system are periodically monitored. Frequencies of testing and control parameters are consistent with the EPRI guidelines for blended glycol formulations. These glycol containing systems are monitoredfor the presenceof microbiologicalactivity in accordancewith the EPRI Closed-Cycle Cooling Water guidelines."

Follow up RAI 3.2.2.2.4.2-1A (also applicable as follow up to RAI 3.3.2.2.2-1)

Background:

By letter dated February 24, 2011, the staff issued RAI 3.2.2.2.4.2-1 concerning the further evaluation for reduction of heat transfer in stainless steel heat exchanger tubes exposed to treated water environment, and requested the technical basis for not managing reduction in heat transfer due to fouling as an aging effect. In its response dated March 22, 2011, the applicant stated that fouling of these components would only occur through the buildup of corrosion products, and since the Seabrook's treated borated water contains boron, a corrosion inhibitor, this was not a credible aging effect/mechanism. The response also stated this determination was based on plant and industry operating experience, in that, fouling has not been identified in treated borated water environment which caused reduction of heat transfer in stainless steel heat exchanger tubes. The response further stated that Seabrook's conclusion is consistent with the NRC staff conclusions as stated in the Beaver Valley (Section 3.2.2.3.2) and Prairie Island (Section 3.2.2.2.4) Safety Evaluation Report (SER).

United States Nuclear Regulatory Commission SBK-L-1 1123 / Enclosure 1 / Page 3 With regard to the cited SERs, the staff notes that in every instance where heat transfer was identified as an intended function in treated borated water for Engineered Safeguards and Auxiliary Systems, both of the associated LRAs had line items that addressed reduction in heat transfer as an aging effect requiring management. The staff also noted for Beaver Valley that in many of these instances, in addition to using the water chemistry AMP, a separate verification of the AMP's effectiveness was also performed.

Issue:

The RAI response stated that reduction in heat transfer is not an aging effect in a treated borated water environment, and stated that this determination was based on plant and industry operating experience. The staff notes that the SRP-LR clearly states that heat transfer functions should be considered for heat exchanger components because heat transfer may be a primary safety function.

Furthermore, Branch Technical Position RLSB-1, for Applicable Aging Effects states that an aging effect should be identified as applicable for license renewal even if there is a prevention or mitigation program associated with that aging effect. The staff noted that Seabrook's LRA cited heat transfer as an intended function for heat exchanger components exposed to treated borated water in the containment building spray, residual heat removal, chemical and volume control, and spent fuel pool cooling systems; however, the LRA did not cite an aging management program to manage reduction of heat transfer.

In addition, the staff noted that all the LRA's submitted for pressurized water reactors (PWRs) in the last three years have included reduction in heat transfer as an aging effect requiring management in treated borated water for heat exchanger components.

Request:

Provide specific technical justification to demonstrate that heat exchanger tubes exposed to treated

• borated water which have an intended function of heat transfer need not inclutde reduction of heat transfer as an aging effect requiring management. As part of the justification, include the plant-specific and industry operating experience cited in the response, showing that reduction in heat transfer had been specifically included as an attribute being investigated, and subsequently demonstrated not to be a credible aging effect/mechanism.

United States Nuclear Regulatory Commission SBK-L- 11123 / Enclosure 1 / Page 4 NextEra Enerev Seabrook ResDonse In response to this RAI, the Water Chemistry Program will be assigned as the aging management program to manage reduction of heat transfer due to fouling in the stainless steel heat exchanger tubes or cooling coils in treated borated water environment. Accordingly, the following changes have been made to the LRA as follows:

1. In Table 3.1.2-1, on page 3.1-46, the following new row is added before the 1 st row as follows:

Heat Exchanger Components Heat Treated (Reactor Transfer Stainless Borated Reduction Water CoolanthPump Steel Water of Heat Chemistry None None H, 4 Thermal Pressure (External) Transfer Program Barrier Heat Boundary Exchanger Cooling Coil)

2. In Table 3.1.2-1, on page 3.1-68, new note 4 is added as a plant specific note as follows:

4 The aging effect/mechanism of reduction of heat transferdue tofouling is not in NUREG-1801for this component, material,and environment combination. The Water Chemistry Program is used to manage the aging effect for this component, material, and environment combination.

3. In Table 3.2.2-2, on page 3.2-47, the following new row is added after the 4th row as follows:

Heat Heat Exchanger Transfer Treated Reduction Water Components Stainless Borated of Heat Chemistry None None H, 4 (1-CBS-E-16 Pressure Steel Water Transfer Program and 16B oundary (Internal)

Tubes) B on pa gs

4. In Table 3.2.2-2, on page 3.2-48, the following new row is added after the 4t1h row as follows:

United States Nuclear Regulatory Commission SBK-L-l 1123 / Enclosure 1 / Page 5

5. In Table 3.2.2-2, on page 3.2-58, new note 4 is added as a plant specific note as follows:

4 The agingeffect/mechanism of reduction of heat transferdue to fouling is not in NUREG-1801for this component, material,and environment combination. The Water Chemistry Program is used to manage the aging effect for this component, material, and environment combination.

6. In Table 3.2.2-3, on page 3.2-61, a new row is added after the 5 th row as follows:

7. In Table 3.2.2-3, on page 3.2-63, a new row is added before the 1 st row as follows:

Heat Heat Exchanger Transfer Treated Reduction Water Components Stainless Borated of Heat Chemistry None None (1-RH-E-188A Pressure Steel Water Transfer Program and 188B Boundary (Internal)

Tubes) Boundary

8. In Table 3.2.2-3, on page 3.2-71, new note 4 is added as a plant specific note as follows (Note 3 was added in response to RAI 3.2.2.3-01 in Letter SBK-L-I 1015, dated 2-3-2011, on page 21 of Enclosure 1):

4 The aging effect/mechanism of reduction of heattransferdue tofouling is not in NUREG-1801for this component, material,and environment combination. The Water Chemistry Program is used to manage the aging effect for this component, material, and environment combination.

9. In Table 3.3.2-3, on page 3.3-145, a new row is added before the 1 st row as follows:

10. In Table 3.3.2-3, on page 3.3-145, a new row is added after the 3 rd row as follows:

United States Nuclear Regulatory Commission SBK-L- 11123 / Enclosure 1 / Page 6

11. In Table 3.3.2-3, on page 3.3-147, a new row is added after the 6 th row as follows:

12. In Table 3.3.2-3, on page 3.3-149, a new row is added after the 7th row as follows:

13. In Table 3.3.2-3, on page 3.3-152, a new row is added before the 1st row as follows:

14. In Table 3.3.2-3, on page 3.3-154, a new row is added after the 1 st row as follows:

15. In Table 3.3.2-3, on page 3.3-154, a new row is added after the 4th row as follows:

16. In Table 3.3.2-3, on page 3.3-156, a new row is added after the 6th row as follows:

United States Nuclear Regulatory Commission SBK-L-1 1123 / Enclosure 1 / Page 7

17. In Table 3.3.2-3, on page 3.3-158, a new row is added after the 611 row as follows:

18. In Table 3.3.2-3, on page 3.3-159, a new row is added after the 1st row as follows:

19. In Table 3.3.2-3, on page 3.3-188, new note 9 is added as a plant specific note as follows (Note 8 was revised in response to RAI 3.2.2.2.6-2 in Letter SBK-L- 11062, dated 4-5-2011, on page 3 of Enclosure 1):

9 The aging effect/mechanism of reduction of heattransferdue tofouling is not in NUREG-1801for this component, material,and environment combination. The Water Chemistry Program is used to manage the aging effect for this component, material, and environment combination.

20. In Table 3.3.2-39, on page 3.3-484, a new row is added after the 7h row as follows:

21. In Table 3.3.2-39, on page 3.3-491, new note 5 is added as a plant specific note as follows:

5 The agingeffect/mechanism of reduction of heattransferdue tofouling is not in NUREG-1801for this component, material,and environment combination. The Water Chemistry Program is used to manage the aging effect for this component, material, and environment combination.

United States Nuclear Regulatory Commission SBK-L- 11123 / Enclosure 1 / Page 8 Request for Additional Information (RAI) 3.3.2.3.4-1 Second Follow up

Background:

By letter dated March 30, 2011, the staff issued RAI 3.3.2.3.4-1, the staff requested that the applicant state the chlorine concentration in the chlorination system and state why no aging effect will occur, or propose an aging management program for the fiberglass components in LRA Table 3.3.2-4. In its response dated April 22, 2011, the applicant stated that components in the chlorination system could be exposed to chlorine levels up to 5400 ppm. The applicant also stated that the components exposed to chlorine are constructed of fiberglass reinforced vinyl ester or bisphenol-A polyester. The applicant further stated that based on input from the vendor of the components, given the system operating parameters, less than 65 'F, pH greater than 10, and no direct ultraviolet exposure; and plant-specific operating experience to date, there is no potential aging effect.

Issue:

Based on independent research, the staff does not agree with the applicant's assessment that there is no aging effect for these components. While the applicant's response to the RAI establishes that the materials are suitable for the design parameters of the system, proper design does not establish the basis for a 60-year life with no aging effects when the environment is an oxidizer and the material is an organic polymer.

Request:

Please state what inspections have been performed to establish a baseline of operating experience and what inspections will be conducted (e.g., quantity, type, frequency, timing) to manage aging of the fiberglass piping and fittings in the chlorination system exposed to raw water, including sodium hypochlorite.

NextEra Energy Seabrook Response:

During Refueling Outage 14 (April 2011), the Chlorination System piping was disassembled during the implementation of an engineering change. The piping was found to be in excellent condition with no signs of aging related degradation.

To validate the suitability of this material in this environment and assure that the components will continue to perform their license renewal intended function during the period of extended operation, the Chlorination System fiberglass reinforced vinyl ester or bisphenol-A polyester pipe within the scope of license renewal will be inspected under the One-Time Inspection Program.

United States Nuclear Regulatory Commission SBK-L- 11123 / Enclosure 1 / Page 9 Based on the above discussion, the following changes have been made to the LRA:

1. In section 3.3.2.1.4, on page 3.3-11, the following aging effect is added to the list of aging effects listed under Aging Effects Requiring Management.

• Cracking,Blistering,Change in Color

2. In section 3.3.2.1.4, on page 3.3-12, the following program is added to the list of aging management programs listed under Aging Management Programs.

• One Time Inspection Program (B.2.1.20)

3. In Table 3.3.2-4, on page 3.3-190, the 1st row is revised as follows:

Piping Leakage Cracking, Raw Water Blistering, One-Time Fittings (Spatial) (Internal) Change in Inspection None None F, 6 Color

4. In Table 3.4.2-3, on page 3.3-194, new Plant Specific Note 6 is added as follows:

6 Aging Effect not in NUREG-1801for this materialand environment combination. The raw water environment downstream of the sodium hypochlorite injection is chlorinated.

Seabrook Station has determined the material chosen for this application,fiberglass reinforcedvinyl ester or bisphenol-A polyester, to be unaffected by long term exposure to this environment. To validate thisposition, the pipingof interest will be inspectedunder the One-Time Inspection Program.

5. The following incorporates a change in Appendix A that was inadvertently omitted in response to RAI 3.4.2.3-01 (SBK-L-11015 dated February 3,2011). In Appendix A, in Section A.2.1.20, on page A-13, in the last paragraph, add a 4th bullet as follows:

• Verification of stainless steel components in the Auxiliary Steam Heating System, to manage the effects of cracking in a steam environment.

6. In Appendix A, in Section A.2.1.20, on page A-13, in the last paragraph, a 5 th bullet is added as follows:

Verification of reinforcedfiberglass vinyl ester or bisphenol-A polyester pipe in the ChlorinationSystem, to manage the aging effects of cracking, blistering,and change in color in chlorinatedraw water environment.

United States Nuclear Regulatory Commission SBK-L- 11123 / Enclosure I / Page 10

7. In Appendix B, in Section B.2.1.20, on page B-1 18, in the first full paragraph, a 5th bullet is added as follows:

• Verification of reinforcedfiberglass vinyl ester or bisphenol-A polyester pipe in the ChlorinationSystem, to manage the aging effects of cracking, blistering,and change in color in chlorinatedraw water environment.

8. In Appendix B, in Section B.2.1.20, on page B-I 18, the last paragraph is revised to read as follows:

This program assesses aging effects of loss of material due to corrosion (general, pitting, crevice, or galvanic); loss of material due to microbiological influenced corrosion; loss of material due to fouling; reduction of heat transfer due to fouling; cracking due to stress corrosion cracking and cyclic loading; and cracking, blistering, change in color of susceptible components within License Renewal scope. This program will select the locations to be inspected, provide the inspection criteria, evaluate the results of the inspections and provide recommendations for additional inspections, as necessary. The results of these inspections will be evaluated for impact throughout the relevant systems at Seabrook Station. They will also determine the need for additional inspections to manage this aging effect.

Seabrook follow-up RAI 4.3.3-1c

Background:

In its response to RAI 4.3.3-lb dated April 22, 2011, the applicant revised LRA Section 4.3.3 indicating that "the effects of fatigue on these limiting locations will be monitored by cycle counting under the Seabrook Station Fatigue Monitoring Program". The applicant also revised the TLAA disposition basis to 10CFR 54(c)(1 )(i), that 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.

Issue:

The staff noted that the aging management program for metal fatigue in LRA Appendix B is called Metal Fatigue of Reactor Coolant Pressure Boundary Program and not Fatigue Monitoring Program, as indicated by the applicant in the revised LRA Section 4.3.3. Furthermore, as indicated in SRP-LR Section 4.3.2.1.1.3, the applicant-proposed aging management program to address metal fatigue of the reactor coolant systems components should-be 10 CFR 54.21(c)(1)(iii) and not 10CFR 54(c)(1)(i), as indicated by the applicant in the revised TLAA disposition. The staff also noted that LRA Section A.2.4.2.2.2 is not updated to reflect the changes in LRA Section 4.3.3 due to the letter dated April 22, 2011.

United States Nuclear Regulatory Commission SBK-L- 11123 / Enclosure 1 / Page 11 Request:

(1) Revise and identify the appropriate metal fatigue aging management program in LRA Section 4.3.3.

(2) Revise the TLAA disposition in LRA Section 4.3.3 and Table 4.1-1.

(3) Provide an updated UFSAR supplement section in LRA Section A.2.4.2.2.2 consistent with the changes in LRA Section 4.3.3.

NextEra Energy Seabrook Response:

(1) LRA section 4.3.3 page 4.3-17 and subsequent response to RAI 4.3.3-lb previously provided in SBK-L- 11069, dated April 22, 2011 (Reference 3) is revised as follows:

Fatigue Fatigue is defined as the structural deterioration that can occur as the result of repeated stress/strain cycles caused by fluctuating loads and temperatures. After repeated cyclic loading of sufficient magnitude, microstructural damage can accumulate, leading to macroscopic crack initiation at the most highly affected locations. Subsequent mechanical or thermal cyclic loading can lead to growth of the initiated crack. Corrosion fatigue is included in the degradation description.

Low-cycle fatigue is defined as cyclic loads that cause significant plastic strain in the highly stressed regions, where the number of applied cycles is increased to the point where the crack eventually initiates. When the cyclic loads are such that significant plastic deformation does not occur in the highly stressed regions, but the loads are of such increased frequency that a fatigue crack eventually initiates, the damage accumulated is said to have been caused by high-cycle fatigue. The aging effects of low-cycle fatigue and high-cycle fatigue are additive. Fatigue crack initiation and growth resistance is governed by a number of material, structural and environmental factors, such as stress range, loading frequency, surface condition and presence of deleterious chemical species. Cracks typically initiate at local geometric stress concentrations, such as notches, surface defects, and structural discontinuities. The aging effect is cracking.

For design purposes, the American Society of Mechanical Engineers (ASME) Code Section III fatigue design procedures use a design fatigue curve that is a plot of alternating stress range (Sa) versus the number of cycles to failure (N). The design fatigue curve is based on the unnotched fatigue properties of the material, modified by reduction factors that account for various geometric and moderate environmental effects.

The fatigue usage factor (U) is defined by Miner's rule as the summation of the damage over the total number of design basis transient types (X), as given by the ratio of expected cycles of that type (ni) to the allowable number of cycles (Ni) for the stress ranges associated with that transient:

United States Nuclear Regulatory Commission SBK-L-1 1123 / Enclosure 1 / Page 12 U I i ni__

j=1 N, For ASME Code design acceptance, the cumulative usage factor (CUF) calculated in this manner cannot exceed unity (1.0) for the design lifetime of the component.

In accordance with the ASME Code at the time, CUF values were not calculated for the vessel internals as part of the original design basis. However, generic analyses of Westinghouse-designed internals components have shown that the fatigue usage factors of the internals components are low and the number of fatigue sensitive locations is limited. In addition, a plant-specific fatigue analyses for the limiting vessel internals locations has been performed for the Seabrook plant during power uprate, and from this analysis, the Seabrook plant limiting fatigue locations are: Lower Support Columns, Core Barrel Nozzle, Lower Core Plate and Upper Core Plate. The effects of fatigue on these limiting locations will be monitored by cycle counting under the Metal Fatigue of Reactor Coolant PressureBoundary Program,B.2.3.1 during the period of extended operation to validate that the number of design cycles assumed in the analyses will not be exceeded.

(2) LRA section 4.3.3 on page 4.3-19 as previously modified in SBK-L-11069, dated April 22, 2011 (Reference 3) is revised as follows:

Disposition Aging Management,10 CFR 54.21(c)(1)(iii) Validati*n, 10 CFR 51.2 1()(1)(i) -The Metal Fatigue of Reactor Coolant Pressure Boundary Program, B.2.3.1 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.

United States Nuclear Regulatory Commission SBK-L-1 1123 /Enclosure 1 /Page 13 LRA Table 4.1-1 on page 4.1-5 and previously modified in SBK-L- 11069, dated April 22, 2011 (Reference 3) is revised as follows:

Table 4.1-1 Time-Limited Aging Analyses Applicable to Seabrook Station LRA Sectio TLAA Disposition Description Method(s)

Category n

2. Metal Fatigue Of Piping And Components 4.3 Nuclear Steam Supply System (NSSS) Pressure Vessel and §54.21(c)(1)(i) 4.3.1 Component Fatigue Analyses Supplementary ASME Section III, Class 1 Piping and §54.21(c)(1)(i) 4.3.2 Component Fatigue Analyses Absence of a TLAA for Thermal Stresses in Piping Connected to Reactor Coolant Systems: NRC Bulletin 88- N/A 4.3.2.1 08 NRC Bulletin 88-11, Pressurizer Surge Line Thermal §54.21(c)(l)(i) 4.3.2.2 Stratification Reactor Vessel Internal Aging Management 4.3.3

§54.21(c)(1) (iii)

Environmentally-Assisted Fatigue Analyses §54.21(c)(1)(ii) 4.3.4

§54.21 (c)(1)(iii)

Steam Generator Tube, Loss of Material and Fatigue from §54.21(c)(1)(i) 4.3.5 Flow-Induced Vibration Absence of TLAAs for Fatigue Crack Growth, Fracture Mechanics Stability, or Corrosion Analyses Supporting N/A 4.3.6 Repair of Alloy 600 Materials Non-Class 1 Component Fatigue Analyses §54.21(c)(1)(i) 4.3.7

United States Nuclear Regulatory Commission SBK-L- 11123 / Enclosure 1 / Page 14 (3) LRA Section A.2.4.2.2.2 as shown on page A-26 is revised as follows:

A.2.4.2.2.2 Reactor Vessel Internals Aging Management The Seabrook Station Reactor Vessel Internals were designed and constructed prior to the development of ASME Code requirements for core support structures. Demonstration that the effects of aging degradation are adequately managed is essential for assuring continued functionality of the reactor internals during the desired plant operating period, including license renewal. The EPRI M.aer.ial Reliabilit.. Pr-egr-afm (....) Reacter inte.. .als vInspe.tion &

Evaluation (!&E) Guidelines (N4R 227) inspcctin rqieents will manage aging cffcets.

The Metal Fatigueof Reactor CoolantPressureBoundaryProgram,B.2.3.1 will monitor the number of design cycles assumed in the fatigue analysis to assure that these will not be exceeded duringthe periodof extended operation. LRA section B.2.1. 7provides the details of the Seabrook Station PWR Vessel InternalsProgram.

In accordance with 10 CFR 54.21(c), the MetalFatigueofReactor CoolantPressureBoundary Program,B.2.3.1 will monitor the number of design cycles assumedin thefatigue analysisto provide assurance that the effects of aging will be adequately managed for the period of extended operation per 10 CFR 54.21(c)(1)(iii).

Enclosure 2 to SBK-L-11123 LRA Appendix A - Final Safety Report Supplement Table A.3 License Renewal Commitment List

United States Nuclear Regulatory Commission SBK-L-1 1123 / Enclosure 2 / Page 2 A.3 LICENSE RENEWAL COMMITMENT LIST UFSAR No. PROGRAM or TOPIC COMMITMENT LOCATION SCHEDULE Program to be implemented prior to the period of extended An inspection plan for Reactor Vessel Internals will be operation. Inspection plan submitted for NRC review and approval, to be submitted to NRC

1. PWR Vessel Internals A.2.1.7 not later than 2 years after receipt of the renewed license or not less than 24 months prior to the period of extended operation, whichever comes first.

Closed-Cycle Cooling Enhance the program to include visual inspection for Prior to the period of

2. CsCe cracking, loss of material and fouling when the in-scope A.2.1.12 extended operation Water

2. systems are opened for maintenance.

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

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

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

4. Load (Related to monitoringA.2.1.13 extended operation Refueling) Handling Systems Enhance the program to include an annual air quality test Prior to the period of Compressed Air requirement for the Diesel Generator compressed air sub A.2.1.14 extee perion

.Monitoring system. extended operation

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

United States Nuclear Regulatory Commission SBK-L- 11123 / Enclosure 2 / Page 3 UFSAR No. PROGRAM or TOPIC COMMITMENT LOCATION SCHEDULE Enhance the program to add inspection requirements such

7. Fire Protection as spalling, and loss of material caused by freeze-thaw, A.2.1.15 Prior to the period of chemical attack, and reaction with aggregates by qualified extended operation.

inspector.

8. Enhance the program to include the performance of visual Prior to the period of Fire Protection inspection of fire-rated doors by a fire protection qualified A.2.1.15 extended operation.

inspector.

Enhance the program to include NFPA 25 guidance for

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

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

10. Enhance the program to include the performance of Prior to the period of Fire Water System periodic flow testing of the fire water system in accordance A.2.1.16 extended operation.

with the guidance of NFPA 25.

Enhance the program to include the performance of periodic visual or volumetric inspection of the internal surface of the fire protection system upon each entry to the system for routine or corrective maintenance. These inspections will be documented and trended to determine if Within ten years prior to

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

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

Enhance the program to include components and aging Prior to the period of

12. Aboveground Steel effects required by the Aboveground Steel Tanks. A.2.1.17 extended operation.

Tanksexeddoeain

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

United States Nuclear Regulatory Commission SBK-L-11123 / Enclosure 2 / Page 4 UFSAR No. PROGRAM or TOPIC COMMITMENT LOCATION SCHEDULE Enhance program to add requirements to 1) sample and

14. analyze new fuel deliveries for biodiesel prior to offloading Prior to the period of Fuel Oil Chemistry to the Auxiliary Boiler fuel oil storage tank and 2) A.2.1.18 extended operation.

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

Enhance the program to add requirements to check for the

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

necessary.

16. Enhance the program to require draining, cleaning and Prior to the period of Fuel Oil Chemistry inspection of the diesel fire pump fuel oil day tanks on a A.2.1.18 extended operation.

frequency of at least once every ten years.

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

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

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

18. Reactor Vessel Enhance the program to specify that all pulled and tested Prior to the period of Surveillance capsules, unless discarded before August 31, 2000, are A.2.1.19 extended operation.

placed in storage.

Enhance the program to specify that if plant operations exceed the limitations or bounds defined by the Reactor

19. Reactor Vessel Vessel Surveillance Program, such as operating at a lower Prior to the period of Surveillance cold leg temperature or higher fluence, the impact of plant A.2.1.19 Prirntotheperiodo operation changes on the extent of Reactor Vessel extended operation.

embrittlement will be evaluated and the NRC will be notified.

United States Nuclear Regulatory Commission SBK-L-1 1123 / Enclosure 2 / Page. 5 UFSAR No. PROGRAM or TOPIC COMMITMENT LOCATION SCHEDULE Enhance the program as necessary to ensure the appropriate withdrawal schedule for capsules remaining in the vessel such that one capsule will be withdrawn at an

20. Reactor Vessel outage in which the capsule receives a neutron fluence that Prior to the period of S.Reianct sel meets the schedule requirements of 10 CFR 50 Appendix H A.2.1.19 extended operation.

Surveillance and ASTM El 85-82 and that bounds the 60-year fluence, and the remaining capsule(s) will be removed from the vessel unless determined to provide meaningful metallurgical data.

Enhance the program to ensure that any capsule removed,

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

including during the period of extended operation.

22. IWithin ten years prior to One-Time Inspection Implement the One Time Inspection Program. A.2.1.20 the period of extended operation.

Implement the Selective Leaching of Materials Program.

23. Selective Leaching of The program will include a one-time inspection of selected Within five years prior to Materials components where selective leaching has not been A.2.1.21 the period of extended identified and periodic inspections of selected components operation.

where selective leaching has been identified.

24. Buried Piping And Tanks Implement the Buried Piping And Tanks Inspection Within ten years prior to Inspection Program. A.2.1.22 entering the period of extended operation One-Time Inspection of Implement the One-Time Inspection of ASME Code Class 1 Within ten years prior to

25. ASME Code Class 1 Implement te oneTme A.2.1.23 the period of extended SSmall Bore-Piping mall Bore-Piping Program. operation.

United States Nuclear Regulatory Commission SBK-L- 11123 / Enclosure 2 / Page 6 UFSAR No. PROGRAM or TOPIC COMMITMENT LOCATION SCHEDULE Enhance the program to specifically address the scope of the program, relevant degradation mechanisms and effects

26. External Surfaces of interest, the refueling outage inspection frequency, the Prior to the period of Monitoring inspections of opportunity for possible corrosion under A.2.1.24 extended operation.

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

Inspection of Internal

27. Surfaces in Implement the Inspection of Internal Surfaces in Prior to the period of Miscellaneous Piping Miscellaneous Piping and Ducting Components Program. A.2.1.25 extended operation.

and Ducting Components

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

changes.

29: Enhance the program to sample the oil for the Switchyard Prior to the period of Lubricating Oil Analysis SF 6 compressors and the Reactor Coolant pump oil A.2.1.26 extended operation.

collection tanks.

Enhance the program to require the performance of a one-

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

period of extended operation.

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

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

requirements.

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

United States Nuclear Regulatory Commission SBK-L- 1 1123 / Enclosure 2 / Page 7 U FSAR No. PROGRAM or TOPIC COMMITMENT LOCATION SCHEDULE Electrical Cables and Connections Not Subject Implement the Electrical Cables and Connections Not

34. to 10 CFR 50.49 pPriorto 10 CFR 50.49 Environmental Qualification period of to theoperation.

Environmental Subject A.2.1.32 extended Qualification Requirements program.

Requirements Electrical Cables and Connections Not Subject

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

Qualification Requirements Used in Instrumentation Circuits program.

Requirements Used in*

Instrumentation Circuits Inaccessible Power Cables Not Subject to 10 Implement the Inaccessible Power Cables Not Subject to Prior to the period of

36. CFR 50.49 10 CFR 50.49 Environmental Qualification Requirements A.2.1.34 Environmental 1ro 50. extended operation.

Qualification program.

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

, extended operation.

38. Prior to the period of

38. Fuse Holders Implement the Fuse Holders program. A.2.1.36 extended operation.

Electrical Cable Connections Not Subject Implement the Electrical Cable Connections Not Subject to

39. to 10 CFR 50.49 10 CFR 50.49 Environmental Qualification Requirements A.2.1.37 Prior to the period of Environmental prga.extended operation.

Qualification program.

Requirements

40. Prior to the period of

40. 345 KV SF 6 Bus Implement the 345 KV SF 6 Bus program. A.2.2.1 extended operation.

United States Nuclear Regulatory Commission SBK-L-I 1123 / Enclosure 2 / Page 8 UFSAR No. PROGRAM or TOPIC COMMITMENT LOCATION SCHEDULE

41. Metal Fatigue of Reactor Enhance the program to include additional transients Prior to the period of Coolant Pressure beyond those defined in the Technical Specifications and A.2.3.1 extended operation.

Boundary UFSAR.

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

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

Boundary components.

The updated analyses will Pressure -Temperature be submitted at the

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

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 Environmentally- identified consists of nickel alloy, the environmentally- At least two years prior to

44. Assisted Fatigue assisted fatigue calculation for nickel alloy will be performed A.2.4.2.3 entering the period of Analyses (TLAA) using the rules of NUREG/CR-6909. extended operation.

(1) Consistent with the Metal Fatigue of Reactor Coolant Pressure Boundary Program Seabrook Station will update the fatigue usage calculations using refined fatigue analyses, if necessary, to determine acceptable CUFs (i.e.,

less than 1.0) when accounting for the effects of the reactor water environment. This includes applying the appropriate Fen factors to valid CUFs determined from an existing fatigue analysis valid for the period of extended operation or from an analysis using an NRC-approved version of the ASME code or NRC-approved alternative (e.g., NRC-

United States Nuclear Regulatory Commission SBK-L- 11123 / Enclosure 2 / Page 9 UFSAR No. PROGRAM or TOPIC COMMITMENT LOCATION SCHEDULE 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).

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

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

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

United States Nuclear Regulatory Commission SBK-L-1 1123 / Enclosure 2 / Page 10 UFSAR No. PROGRAM or TOPIC COMMITMENT LOCATION SCHEDULE Protective Coating Enhance the program to include a review of the previous A.2.1.38 Prior to the period of

48. Monitoring and two monitoring reports. extended operation Maintenance Protective Coating Enhance the program to require that the inspection report is Prior to the period of

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

ASME Section XI, Perform UT testing of the containment liner plate in the A.2.1.27 No later than December 50.Subsection IWE vicinity of the moisture barrier for loss of material. 31, 2015 and repeated at intervals of no more than five refueling outages ASME Section Xl, Perform confirmatory testing and evaluation of the A212 Prior to the period of

51. Subsection IWL Containment Structure concrete ... 8 extended operation ASME Section XI, Implement measures to maintain the exterior surface of the

52. Subsection IWL Containment Structure, from elevation -30 feet to +20 feet, A.2.1.28 By 2013 in a dewatered state.

Reactor Head Closure Replace the spare reactor head closure stud(s) Prior to the period of

53. Studs manufactured from the bar that has a yield strength > 150 A.2.1.3 extended operation.

ksi with ones that do not exceed 150 ksi.

Unless an alternate repair criteria changing the ASME code boundary is permanently approved by the NRC, or the Seabrook Station steam generators are changed to Program to be submitted Steam Generator Tube eliminate PWSCC-susceptible tube-to-tubesheet welds, A.2.1.10 to NRC at least 24 Integrity submit a plant-specific aging management program to months prior to the period manage the potential aging effect of cracking due to of extended operation.

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

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

55. Integrity generator assembly. to assess the condition of the divider plate A.2.1.10 period of extended operation Closed-Cycle Cooling Revise the station program documents to reflect the EPRI Prior to entering the

56. Cloed-Cylem Guideline operating ranges and Action Level values for A.2.1.12 period of extended Water System hydrazine and sulfates. operation.

United States Nuclear Regulatory Commission SBK-L-I 11123 / Enclosure 2 / Page 11 UFSAR No. PROGRAM or TOPIC COMMITMENT LOCATION SCHEDULE Closed-Cycle Cooling Revise the station program documents to reflect the EPRI Prior to entering the Closaed-Cylem,CGuideline operating ranges and Action Level values for A.2.1.12 period of extended Water System Diesel Generator Cooling Water Jacket pH. operation.

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

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

and ASTM D4057-95 required by the GALL XI.M30 Rev 1 Nickel Alloy Nozzles and The Nickel Alloy Aging Nozzles and Penetrations program Prior to the period of Penetrations

59. will implement applicable Bulletins, Generic Letters, and A.2.2.3 extended operation.

staff accepted industry guidelines. extededperaion Buried Piping and Tanks Implement the design change replacing the buried Auxiliary *Prior to entering the

60. Inspection Boilerleak with supply piping capability, indication with a pipe-within-pipe configuration A.2.1.22 period of extended operation.

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

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

Enhance the program to include a statement that sampling -Prior to entering the

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

Ensure that the quarterly CVCS Charging Pump testing is continued during the PEO. Additionally, add a precaution to

63. Flow Induced Erosion the test procedure to state that an increase in the CVCS N/A Prior to the period of Charging Pump mini flow above the acceptance criteria extended operation may be indicative of erosion of the mini flow orifice as described in LER 50-275/94-023.

Soil analysis shall be performed prior to entering the period of extended operation to determine the corrosivity of the soil Prior to entering the

64. Buried Piping and Tanks in the vicinity of non-cathodically protected steel pipe within A.2.1.22 period of extended Inspection the scope of this program. If the initial analysis shows the operation.

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

Implement measures to ensure that the movable incore Prior to entering the

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