SBK-L-18151, Supplement 61 - Additional Changes to the NextEra Energy Seabrook License Renewal Application

Supplement 61 - Additional Changes to the NextEra Energy Seabrook License Renewal Application

ML18241A324
Person / Time
Site:	Seabrook
Issue date:	08/29/2018
From:	Mccartney E NextEra Energy Seabrook
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SBK-L-18151
Download: ML18241A324 (28)

Text

NEXTera ENERGY .

SEABROOK August 29, 2018 10 CFR 54 Docket No. 50-443 SBK-L-18151 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Wash ington, DC 20555-0001 Seabrook Station Supplement 61 - Additional Changes to the NextEra Energy Seabrook License Renewal Application

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-13084, "Seabrook Station NextEra Energy Seabrook License Renewal Application Revised Response to RAI B.2.1.3-5 and RAI 4.7.2-1" May 8, 2013. (Accession Number ML13135A005)

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.

In Reference 2, NextEra Energy Seabrook provided a revised response to RAI B.2.1.3-5 and 4.7 .2-1 concerning fracture mechanics evaluation for performing fatigue assessments, and Reactor Coolant Pump Flywheel examination frequency changes.

Upon discussion with the NRC staff, NextEra Energy Seabrook agreed to supplement a portion of the previously provided information. Enclosure 1 provides additional changes to the License Renewal Application (LRA), Section A.2.4.5 .1 and Section 4.7.2, pertaining to Reactor Coolant Pump Flywheel inspection frequencies. Changes are also provided to the LRA, Section 4.2.1, Section 4.2.4, and Section A.2 .1.4, pertaining to Reactor Vessel Pressure-Temperature Limits and Neutron Fluence Analyses.

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

U.S. Nuclear Regulatory Commission SBK-L-18151IPage2 To facilitate understanding, the changes are explained, and where appropriate, portions of the LRA are repeated with the change highlighted by strikethroughs for deleted text and balded italics for inserted text.

There is one completed regulatory commitment contained in this letter, Commitment 43.

Enclosure 2 provides the revised LRA Appendix A - Final Safety Report Supplement Table A.3, License Renewal Commitment List.

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

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

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

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

Executed on August 29, 2018 Sincerely, NextEra Energy Seabrook, LLC

'--Bic McartneY Regional Vice President - Northern Region Enclosure 1: Additional Changes to the NextEra Energy Seabrook License Renewal Application Enclosure 2: LRA Appendix A - Final Safety Report Supplement Table A.3, License Renewal Commitment List Updated to Reflect Changes to Date cc: D. H. Dorman NRC Region I Administrator J.C. Poole NRC Project Manager P. C. Cataldo NRC Senior Resident Inspector E. H. Gettys NRC Project Manager, License Renewal T. M. Tran NRC Project Manager, License Renewal B. H. Rogers NRC Project Manager, License Renewal A Hiser NRC Staff

U.S. Nuclear Regulatory Commission SBK-L-18151/Page3 B. Rogers NRC Staff A. Billoch NRC Staff

U.S. Nuclear Regulatory Commission SBK-L-18151IPage4 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 perry.plummer@dos.nh.gov Mr. John Giarrusso, Jr., Nuclear Preparedness Manager The Commonwealth of Massachusetts Emergency Management Agency 400 Worcester Road Framingham, MA 01702-5399 John.Giarrusso@massmail.state.ma.us

Enclosure 1 to SBK-L-18151 AddWonalChangestothe NextEra Energy Seabrook License Renewal Application

U.S. Nuclear Regulatory Commission SBK-L-18151/Enclosure1/Page2 License Amendment 134 extended the Reactor Coolant Pump flywheel examination frequency from 10 years to an interval not to exceed 20 years. The following are the resulting License Renewal Application (LRA) changes:

4.7.2 REACTOR COOLANT PUMP FLYWHEEL FATIGUE CRACK GROWTH ANALYSES Disposition Validation, 10 CFR 54.21(c)(1)(i) - Since the number of analyzed start/stop cycles exceeds the 60-year cycle projections, the reactor coolant pump flywheel analysis remains valid for the period of extended operation.

Based on WCAP 15666; Amendment 134 to the Facility Operating License extended the Reactor Coolant Pump (RCP) flywheel examination frequency from a 10 year inspection interval to an interval not to exceed 20 years. During the period of extended operation the reactor coolant pump flywheels will be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision I, August 1975. In lieu of Position C.4.b(I) and C.4.b(2}, this inspection shall be by either of the following examinations:

a. An in-place examination, utilizing ultrasonic testing, over the volume from the inner bore of the flywheel to the circle of one-half the outer radius; or

b. A surface examination, utilizing magnetic particle testing and/or penetrant testing, of the exposed surfaces of the disassembled flywheel.

A.2.4.5.1 Reactor Coolant Pump Flywheel Fatigue Crack Growth Analyses Westinghouse Report WCAP-14535-A, Rev. 0, "Topical Report on Reactor Coolant Pump Flywheel Inspection Elimination" includes a fatigue crack growth analysis that has been identified as a TLAA. The report was submitted for NRC review and the NRC issued a Safety Evaluation Report in September 1996. The purpose of the report was to provide an engineering basis for elimination of reactor coolant pump (RCP) flywheel in-service inspection requirements for all operating Westinghouse plants and certain Babcock and Wilcox plants. The number of cycles (pump starts and stops) used in this report was 6,000 for a 60-year plant life. Crack growth was shown to be negligible from exposure to these 6,000 cycles.

U.S. Nuclear Regulatory Commission SBK-L-18151IEnclosure1IPage3 Based on WCAP 15666; Amendment 134 to the Facility Operating License extended the reactor coolant pump (RCP) flywheel examination frequency from a 10 year inspection interval to an interval not to exceed 20 years. During the period of extended operation the reactor coolant pump flywheels will be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision I, August 1975. In lieu of Position C.4.b(I) and C.4.b(2), this inspection shall be by either of the following examinations:

a. An inap/ace examination, utilizing ultrasonic testing, over the volume from the inner bore of the flywheel to the circle of one-half the outer radius; or

b. A surface examination, utilizing magnetic particle testing and/or penetrant testing, of the exposed surfaces of the disassembled flywheel.

Based on the current cycle count projected to 60 years, the projected cycle count is much less than the analyzed cycle counts of 6,000 cycles. The reactor coolant pump flywheel analysis remains valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i).

License Amendment 151 approved 10CFR50, Appendix G, Pressure-Temperature Limits applicable to 55 Effective Full Power Years (EFPY). As a result, the following LRA changes have been added as the last paragraphs in the Analysis portion of Section 4.2.1, Neutron Fluence Analysis:

The 60-Year Neutron Fluence Projections (55 EFPY) were subsequently updated in support of the preparation of the 55 EFPY pressure temperature limit curve TLAA in Section 4.2.4. The updated neutron fluence analysis was based on nuclear cross-section data derived from ENDFIB-Vl.3 and made use of the latest available calculation tools. Furthermore, the neutron transport evaluation methodologies follow the guidance of NRG Regulatory Guide 1.190. Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRG-approved methodology described in WCAP-14040-A, Revision 4 [WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Coo/down Limit Curves," J. D. Andrachek, et al., May 2004].

The updated 55 EFPYprojected f/uence values are reported in WCAP-17441-NP, Rev 0 and submitted as part the technical basis for License Amendment Request 14-04 to revise the Reactor Vessel Pressure-Temperature Limits applicable for 55

U.S. Nuclear Regulatory Commission SBK-L-18151IEnclosure1IPage4 EFPY and subsequently approved in License Amendment 151(ML15096A255).

The updated 55 EFPY projected f/uence values were compared to the projected fluence values shown in Table 4.2.1-1, above, and the values in Table 4.2.1-1 were found to be bounding. Therefore the neutron f/uence and projected values used to evaluate the other TLAA 's in subsection 4.2.2 for Upper Shelf Energy Analyses and subsection 4.2.3 for Pressurized Thermal Shock Analyses are bounding.

License Amendment 151 approved 10CFR50, Appendix G, Pressure-Temperature Limits applicable to 55 Effective Full Power Years (EFPY). As a result, the following LRA changes have been added as the first paragraph in the Analysis portion of Section 4.2.4, as well as the Disposition portion of Section 4.2.4, Reactor Vessel Pressure-Temperature Limits, Including Low Temperature Overpressure Protection Limit:

4.2.4 REACTOR VESSEL PRESSURE-TEMPERATURE LIMITS, INCLUDING LOW TEMPERATURE OVERPRESSURE PROTECTION LIMIT Analysis The provisions of 10 CFR 50, Appendix G, require Seabrook to operate within the currently licensed P-T limit curves. These curves are required to be maintained and updated as necessary to maintain plant operation consistent with 10 CFR 50. The Reactor Vessel Integrity Surveillance Program maintains the P-T limit curves for the period of extended operation. Prior to the period of extended operation, updated P-T limit calculations will be prepared using fluence values valid for the Seabrook Station reactor vessel beltline region materials, inlet and outlet nozzles, and closure head flange locations for normal heatup, normal cooldown, and in-service leak and hydrostatic test conditions. The current heatup and cooldown limit curves are valid for 20 EFPY. In determining the allowable operating pressure-temperature limits, the minimum bolt-up temperatures, minimum temperature of core criticality, pressure test limits and low-temperature overpressure protection (LTOP) system limits are determined. These P-T limits are expressed in the form of a set of curves of allowable pressure versus temperature (P-T limit curves). These curves are updated on a periodic basis to account for increasing vessel fluence.

Disposition Aging Management, 10 CFR 54.21(c)(1)(iii)- The provisions of 10 CFR 50, Appendix G, require Seabrook to operate within the currently licensed P-T limit curves. These curves are required to be maintained and updated as necessary to maintain plant operation consistent with 10 CFR 50. The Reactor Vessel Integrity

U.S. Nuclear Regulatory Commission SBK-L-18151/Enclosure1 /Page5 Surveillance Program maintains the P-T limit curves for the period of extended operation. Therefore, the P-T limit curves TLAA has been dispositioned in accordance with 10 CFR 54.21(c)(1)(iii). Prior to the period of extended operation, updated P T limit calculations \Nill be prepared for the Seabrook Station reactor vessel beltline region materials, inlet and outlet nozzles, and closure head flange locations for normal heatup, normal cooldown, and in service leak and hydrostatic test conditions. Subsequent to the License Renewal Application, License Amendment Request 14-04 to revise the Reactor Vessel Pressure-Temperature Limits applicable for 55 EFPY was submitted using updated fluence projections identified in subsection 4.2.1, updated limiting 114T and 3/4T limiting ART values, and approved in Technical Specification Amendment 151 (ML15096A255).

The Reactor Vessel Integrity Surveillance Program, B.2.1.19 monitors reactor vessel embrittlement. This program provides data to update the P-T limits and, therefore, permits Seabrook Station to manage the P-T limits going forward in accordance with 10 CFR 54(c)(1 )(iii). Seabrook Station 1.vill submit updates to the PT curves and LTOP limits to the NRG at the appropriate time to comply with 10 CFR 50 Appendix G.

A.2.4.1.4 Reactor Vessel Pressure-Temperature Limits, Including Low Temperature Overpressure Protection Limits Title 10 CFR Part 50, Appendix G requires that the reactor pressure vessel be maintained within established pressure-temperature (P-T) limits, including heatup and cooldown operations. The P-T limits must account for the anticipated reactor vessel fluence. The current minimum Lovv Temperature Overpressure Protection (LTOP) setpoint for Seabrook Station is 561 psig. The current Low Temperature Overpressure Protection (L TOP) system uses a combination of residual heat removal suction relief valves and/or power operated relief valves as identified in Technical Specifications.

The current Seabrook Station P-T and Low Temperature Overpressure Protection (LTOP) limit calculations are effective through ~ 55 EFPY. The 55 EFPY P-T curves and L TOP limits meet the criteria of ASME Code, Section XI, Appendix G, and are in compliance with the fracture toughness requirements of 10 CFR 50.60 and 10 CFR 50, Appendix G through 55 EFPY. Heatup and cooldovm PT limit curves for 55 EFPY V.'ill be prepared using the most limiting value of RTNW (reference nil ductility transition temperature) corresponding to the limiting material in the beltline region of the reactor vessel. This is determined by using the unirradiated reactor vessel material fracture toughness properties adjusted to account for the estimated irradiation induced shift in the Reference Temperature Nil Ductility Transition (ARTNG+}.-

The P T and LTOP limit analyses will not be submitted at this time. The effects of aging on the intended function(s) will be adequately managed for the period of

U.S. Nuclear Regulatory Commission SBK-L-18151IEnclosure1IPage6 extended operation in accordance with 10 CFR 54(c)(1 )(iii).~eabrool< Station will submit updates to the P T curves and LTOP limits to the NRG at the appropriate time to comply *.vith 10 CFR 50 /\ppendix G.

Based on the issuance of Amendment No. 151, License Renewal Commitment No. 43 is now complete and the revised A.3 table is shown below:

+he Hf)EiateEI analyses ,,,,,ill ee Pressure -

s1:1emitteEI at the Temperature Seabrook Station will submit updates to the P-T a13f)f8f)riate time te Limits, including curves and LTOP limits to the NRC at the 43 A.2.4.1.4 eemf)ly witfi IQ GFR Low Temperature appropriate time to comply with I 0 CFR 50

~Q Af)f)ettEii~< G, Overpressure Appendix G.

Fraet1:1re +e1:1glrness Protection Limits Req1:1irements.

Complete

Enclosure 2 to SBK-L-18151 LRA Appendix A - Final Safety Analysis Report Supplement Table A.3, License Renewal Commitment List Updated to Reflect Changes to Date

U.S. Nuclear Regulatory Commission SBK-L-18151IEnclosure2 I Page 2 A.3 LICENSE RENEWAL COMMITMENT LIST UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Provide confirmation and acceptability of the implementation ofMRP-

1. PWR Vessel Internals 227-A by addressing the plant-specific Applicant/Licensee Action Items A.2.1.7 Complete outlined in section 4.2 of the NRC SER.

Enhance the program to include visual inspection for cracking, loss of Prior to the period of extended

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

maintenance.

Inspection of Overhead Heavy Enhance the program to monitor general corrosion on the crane and Prior to the period of extended

3. Load and Light Load (Related to trolley structural components and the effects of wear on the rails in the A.2.1.13 operation.

Refueling) Handling Systems rail system.

Inspection of Overhead Heavy Prior to the period of extended

4. Load and Light Load (Related to Enhance the program to list additional cranes for monitoring. A.2.1.13 operation.

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

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

Enhance the program to perform visual inspection of penetration seals by Prior to the period of extended

6. Fire Protection A.2.l.15 a fire protection qualified inspector. operation.

Enhance the program to add inspection requirements such as spalling, Prior to the period of extended

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

with aggregates by qualified inspector.

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

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

U.S. Nuclear Regulatory Commission SBK-L-18151IEnclosure2 I Page 3 Enhance the program to include NFPA 25 (2011 Edition) guidance for "where sprinklers have been in place for 50 years, they shall be replaced Prior to the period of extended

9. Fire Water System A.2.1.16 or representative samples from one or more sample areas shall be operation.

submitted to a recognized testing laboratory for field service testing".

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

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

(2011 Edition).

Enhance the program to include the performance of periodic visual or volumetric inspection of the internal surface of the fire protection system upon each entry to the system for routine or corrective maintenance to evaluate wall thickness and inner diameter of the fire protection piping ensuring that corrosion product buildup will not result in flow blockage due to fouling. Where surface irregularities are detected, follow-up volumetric examinations are performed. These inspections will be Within ten years prior to the

11. Fire Water System A.2.1.16 documented and trended to determine if a representative number of period of extended operation.

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, 3) Visual, surface, and volumetric examinations of the Within I 0 years prior to the

12. Aboveground Steel Tanks outside and inside surfaces for managing the aging effects of loss of A.2.1.17 period of extended operation.

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

U.S. Nuclear Regulatory Commission SBK-L-18151IEnclosure2 I Page 4 Enhance the program to perform exterior inspection of the fire water storage tanks annually for signs of degradation and include an ultrasonic Within ten years prior to the

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

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

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

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

Boiler 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 A.2.1.18 operation.

quarter 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 A.2.1.18 operation.

ten years.

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

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

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

Enhance the program to specify that all pulled and tested capsules, unless Prior to the period of extended

18. Reactor Vessel Surveillance A.2.1.19 discarded before August 31, 2000, are placed in storage. operation.

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

19. Reactor Vessel Surveillance Program, such as operating at a lower cold leg temperature or higher A.2.1.19 operation.

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

U.S. Nuclear Regulatory Commission SBK-L-18151IEnclosure2 I Page 5 Enhance the program as necessary to ensure the appropriate withdrawal schedule for capsules remaining in the vessel such that one capsule will be withdrawn at an outage in which the capsule receives a neutron Prior to the period of extended

20. Reactor Vessel Surveillance fluence that meets the schedule requirements of 10 CFR 50 Appendix H A.2.1.19 operation.

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, without the intent to test it, is stored in a manner which maintains it in a condition Prior to the period of extended

21. Reactor Vessel Surveillance A.2.1.19 which would permit its future use, including during the period of operation.

extended operation.

Within ten years prior to the

22. One-Time Inspection Implement the One Time Inspection Program. A.2.1.20 period of extended operation.

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

23. Selective Leaching of Materials A.2.1.21 selective leaching has not been identified and periodic inspections of period of extended operation.

selected components where selective leaching has been identified.

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

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

25. A.2.1.23 Code Class l Small Bore-Piping Piping Program. period of extended operation.

Enhance the program to specifically address the scope of the program, relevant degradation mechanisms and effects of interest, the refueling Prior to the period of extended

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

the required periodic reviews to determine program effectiveness.

U.S. Nuclear Regulatory Commission SBK-L-18151 I Enclosure 2 I Page 6 Inspection oflnternal Surfaces in Implement the Inspection of Internal Surfaces in Miscellaneous Piping Prior to the period of extended

27. Miscellaneous Piping and Ducting A.2.l.25 and Ducting Components Program. operation.

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

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

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

29. Lubricating Oil Analysis A.2.1.26 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, Subsection Prior to the period of extended

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

Enhance procedure to add the aging effects, additional locations, Prior to the period of extended

32. Structures Monitoring Program A.2.1.31 inspection frequency and ultrasonic test requirements. operation.

Enhance procedure to include inspection of opportunity when planning Prior to the period of extended

33. Structures Monitoring Program A.2.1.31 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 IO CFR Prior to the period of extended

34. A.2.1.32 Environmental Qualification 50.49 Environmental Qualification Requirements program. operation.

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

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

Requirements Used in Instrumentation Circuits program.

Instrumentation Circuits

U.S. Nuclear Regulatory Commission SBK-L-18151IEnclosure2 I Page 7 Inaccessible Power Cables Not Subject to 10 CFR 50.49 Implement the Inaccessible Power Cables Not Subject to I 0 CFR 50.49 Prior to the period of extended

36. A.2.1.34 Environmental Qualification Environmental Qualification Requirements program. operation.

Requirements Prior to the period of extended

37. Metal Enclosed Bus Implement the Metal Enclosed Bus program. A.2.1.35 operation.

Prior to the period of extended

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

Electrical Cable Connections Not Subject to I 0 CFR 50.49 Implement the Electrical Cable Connections Not Subject to 10 CFR Prior to the period of extended

39. A.2.1.37 Environmental Qualification 50.49 Environmental Qualification Requirements program. operation.

Requirements Prior to the period of extended

40. 345 KV SF6 Bus Implement the 345 KV SF6 Bus program. A.2.2.1 operation.

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

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

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

42. A.2.3.1 Pressure Boundary transients to monitor cumulative usage on selected components. operation.

i:fie Hpaa-teEl fffialyses 'i'<'ill ae Pressure -Temperature Limits, Seabrook Station will submit updates to the P-T curves and LTOP limits sHamittea a-t tfie aflpFepFia-te time

43. including Low Temperature to the NRC at the appropriate time to comply with 10 CFR 50 Appendix A.2.4.1.4 te eemply' witfi l G GFR: §Q Overpressure Protection Limits G. Appenaix G, Fmerurn :i:eHgfiness R:eEjl:!ifements. Complete

U.S. Nuclear Regulatory Commission SBK-L-18151IEnclosure2 I 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 Environmentally-Assisted Fatigue the reactor water environment. This includes applying the appropriate At least two years prior to the

44. A.2.4.2.3 Analyses (TLAA) Fen factors to valid CUFs determined from an existing fatigue analysis period of extended operation.

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-approved code case).

(2) If acceptable CUFs cannot be demonstrated for all the selected locations, then additional plant-specific locations will be evaluated. For the additional plant-specific locations, ifCUF, including environmental effects is greater than 1.0, then Corrective Actions will be initiated, in accordance with the Metal Fatigue of Reactor Coolant Pressure Boundary Program, B.2.3 .1. Corrective Actions will include inspection, repair, or replacement of the affected locations before exceeding a CUF of 1.0 or the effects of fatigue will be managed by an inspection program that has been reviewed and approved by the NRC (e.g., periodic non-destructive examination of the affected locations at inspection intervals to be determined by a method accepted by the NRC).

U.S. Nuclear Regulatory Commission SBK-L-18151IEnclosure2 I Page 9 NextEra will obtain additional cores in the vicinity of20% of the extensometers and perform modulus testing. Using these test results, At least 5 years prior to the period Alkali-Silica Reaction (ASR) NextEra will determine the change in through-thickness expansion since of extended operation (initial

45. A.2.1.31.A Monitoring Program installation of the extensometers and compare it to change determined study) and 10 years thereafter from extensometer readings. Consistency between these results will (follow-up study).

provide additional corroboration of the methodology in MPR-4153.

Protective Coating Monitoring and Enhance the program by designating and qualifying an Inspector Prior to the period of extended

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

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

47. A.2.1.38 Maintenance marker pen, mirror, measuring tape, magnifier, binoculars, camera with operation.

or without wide angle lens, and self sealing polyethylene sample bags."

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

48. A.2.1.38 Maintenance reports. operation.

Enhance the program to require that the inspection report is to be Protective Coating Monitoring and evaluated by the responsible evaluation personnel, who is to prepare a Prior to the period of extended

49. A.2.1.38 Maintenance summary of findings and recommendations for future surveillance or operation.

repair.

ructural deformation will also eted during OR16. Repeat Perform UT of the accessible areas of the containment liner plate in the ASME Section XI, Subsection containment liner UT thickness

50. vicinity of the moisture barrier for loss of material. Perform opportunistic A.2.1.27 IWE examinations at intervals of no UT of inaccessible areas.

more than five (5) refueling outages.

U.S. Nuclear Regulatory Commission SBK-L-18151 I Enclosure 2 I Page 10 Enhance the program to manage the aging effects for closure bolting within air and gas filled systems by using an applicable inspection technique that ensures the integrity of bolted joints will be demonstrated.

For closure bolting within systems at atmospheric pressure, tightness Prior to the period of extended

51. Bolting Integrity checks will be performed on 20 percent of bolts with a maximum of25 A.2.1.9 operation.

bolts per population. Populations will be of the same material and environment combination. Inspections will occur before the period of extended operation, and then every l 0 years after the initial inspection date.

ASME Section XI, Subsection Implement measures to maintain the exterior surface of the Containment

52. A.2.1.28 Complete IWL Structure, from elevation -30 feet to +20 feet, in a dewatered state.

Replace the spare reactor head closure stud(s) manufactured from the bar Prior to the period of extended

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

U.S. Nuclear Regulatory Commission SBK-L-18151IEnclosure2 I Page 11 NextEra will address the potential for cracking of the primary to secondary pressure boundary due to PWSCC oftube-to-tubesheet welds using one of the following two options:

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

54. Steam Generator Tube Integrity A.2.1.10 Complete life of 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 oftube-to-tubesheet weld cracking, or redefining the pressure boundary in which the tube-to-tubesheet weld is no longer included and, therefore, is not required for reactor coolant pressure boundary function. The redefinition of the reactor coolant pressure boundary must be approved by the NRC as part of a license amendment request.

Concrete Surface Suspect areas identified during the 2010 and 2016 ASME Section XI, Subsection

55. Containment IWL inspections will be incorporated into the Seabrook A.2.l.28 September I, 2020 IWL Station Containment Inservice Inspection (CISI) Plan.

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

56. A.2.1.12 System operating ranges and Action Level values for hydrazine and sulfates. operation.

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

57. operating ranges and Action Level values for Diesel Generator Cooling A.2.1.12 System 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 I

U.S. Nuclear Regulatory Commission SBK-L-18151IEnclosure2 I Page 12 The Nickel Alloy Aging Nozzles and Penetrations program will Nickel Alloy Nozzles and Prior to the period of extended

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

industry guidelines.

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

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

capability.

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

61. A.2.1.14 Program compressors on a frequency of every 10 years. period of extended operation.

Enhance the program to include a statement that sampling frequencies Prior to the period of extended

62. Water Chemistry A.2.1.2 are 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 Prior to the period of extended

63. Flow Induced Erosion state that an increase in the CVCS Charging Pump mini flow above the A.2.1.2 operation.

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

Soil analysis shall be performed prior to entering the period of extended operation to determine the corrosivity of the soil in the vicinity of non-Within ten years prior to the

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

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

Implement measures to ensure that the movable incore detectors are not

65. Flux Thimble Tube Complete.

returned to service during the period of extended operation. NIA

U.S. Nuclear Regulatory Commission SBK-L-18151IEnclosure2 I Page 13 NextEra will perform an integrated review of expansion trends at Seabrook Station by conducting a periodic assessment of ASR expansion behavior to confirm that the MPR/FSEL large-scale test programs remain applicable to plant structures. This review will include the following specific considerations:

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

At least 5 years prior to the period Alkali-Silica Reaction (ASR)

66. A.2.1.31.A of extended operation and every Monitoring Program

• Comparison of in-plane expansion to through-thickness expansion 10 years thereafter.

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

• Comparison of in-plane expansions, volumetric expansions, and through-thickness expansions recorded to date to the limits from the MPR/FSEL large-scale test programs and check of margin for future expansion.

Perform one shallow core bore in an area that was continuously wetted from borated water to be examined for concrete degradation and also

67. Structures Monitoring Program expose rebar to detect any degradation such as loss of material. The A.2.1.31 Complete removed core will also be subjected to petrographic examination for concrete degradation due to ASR per ASTM Standard Practice C856.

Perform sampling at the leak off collection points for chlorides, sulfates,

68. Structures Monitoring Program A.2.1.31 Complete pH and iron once every three months.

Replace the Diesel Generator Heat Exchanger Plastisol PVC lined

69. Open-Cycle Cooling Water System A.2.1.11 Complete Service Water piping with piping fabricated from AL6XN material.

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

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

cooling water system.

U.S. Nuclear Regulatory Commission SBK-L-18151/Enclosure21Page14 NextEra has completed testing at the University of Texas Ferguson Structural Engineering Laboratory which demonstrates the parameters Alkali-Silica Reaction (ASR) being monitored and acceptance criteria used are appropriate to manage A.2.l.31A Prior to the period of extended Monitoring Program I Building the effects of ASR.

71. operation.

Deformation Monitoring Program A.2.l.31B NextEra Implement the Alkali-Silica Reaction (ASR) Monitoring Program and Building Deformation Monitoring Program described in B.2.1.3 IA and B.2.1.3 IB of the License Renewal Application.

Enhance the program to include management of wall thinning caused by Prior to the period of extended

72. Flow-Accelerated Corrosion A.2.1.8 mechanisms other than F AC. operation.

Enhance the program to include performance of focused examinations to Inspection oflnternal Surfaces in provide a representative sample of20%, or a maximum of25, of each

73. Miscellaneous Piping and Ducting A.2.1.25 Prior to the period of extended identified material, environment, and aging effect combinations during Components operation.

each I 0 year period in the period of 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 operation.

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 NFP A 25 (2011 Edition) and b) If the presence of sufficient foreign organic or inorganic material Prior to the period of extended

75. Fire Water System A.2.1.16 to obstruct pipe or sprinklers is detected during pipe inspections, the operation.

material will be removed and its source is determined and corrected.

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

U.S. Nuclear Regulatory Commission SBK-L-18151IEnclosure2 I Page 15 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

• main drain tests A.2.1.16 operation.

• 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

• The requirement that ifthere 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 operation.

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

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

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

78. External Surfaces Monitoring A.2.1.24 insulated components exposed to condensation (due to the in-scope operation.

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

Enhance the program to include visual inspection of internal Within 10 years prior to the

79. Open-Cycle Cooling Water System coatings/linings for loss of coating integrity. A.2.1.11 period of extended operation.

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

80. Fire Water System A.2.1.16 period of extended operation.

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

81. Fuel Oil Chemistry A.2.1.18 period of extended operation.

U.S. Nuclear Regulatory Commission SBK-L-18151IEnclosure2 I Page 16 Inspection of Internal Surfaces in Enhance the program to include visual inspection of internal Within 10 years prior to the

82. Miscellaneous Piping and Ducting coatings/linings for loss of coating integrity. A.2.1.25 period of extended operation.

Components Enhance the ASR AMP to install extensometers in all Tier 3 areas of two dimensional reinforced structures to monitor expansion due to

83. Alkali-Silica Reaction Monitoring alkali-silica reaction in the out-of-plane direction. A.2.l.31A Complete Monitoring expansion in the out-of-plane direction will commence upon installation of the extensometers and continue on a six month frequency through the period of extended operation.

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

84. A.2.1.28 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 ofNFPA 25 (2011

85. Prior to the period of extended 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."

86. Enhance the program to include disassembly, inspection, and cleaning of Prior to the period of extended Fire Water System A.2.1.16 the mainline strainers every 5 years. operation.

Increase the frequency of the Open Head Spray Nozzle Air Flow Test

87. Prior to the period of extended Fire Water System from every 3 years to every refueling outage to be consistent with LR- A.2.1.16 operation.

ISG-2012-02, 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

88. Within five years prior to the Fire Water System drains and associated piping are configured to completely drain the A.2.1.16 period of extended operation.

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

U.S. Nuclear Regulatory Commission SBK-L-18151/Enclosure2/Page17 Incorporate Coating Service Level III requirements into the RCP Motor Inspection oflnternal Surfaces in Refurbishment Specification for the internal painting of the motor upper Prior to the period of extended

89. Miscellaneous Piping and Ducting bearing coolers and motor air coolers. All four RCP motors will be A.2.1.25 operation.

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

Implement the PWR Vessel Internals Program. The program will be implemented in accordance with MRP-227-A (Pressurized Water Reactor Prior to the period of extended PWR Vessel Internals A.2.1.7

90. Internals Inspection and Evaluation Guidelines) and NEI 03-08 operation (Guideline for the Management of Materials Issues).

U.S. Nuclear Regulatory Commission SBK-L-18151IEnclosure2 I Page 18 Implement the Building Deformation Monitoring Program Enhance the Structures Monitoring Program to require structural evaluations be performed on buildings and components affected by deformation as necessary to ensure that the structural function is maintained. Evaluations of structures will validate structural performance against the design basis, and may use results from the large-scale test programs, as appropriate. Evaluations for structural deformation will also consider the impact to functionality of affected systems and components (e.g., conduit expansion joints). NextEra will evaluate the specific circumstances against the design basis of the affected system or component.

Enhance the Building Deformation AMP to include additional 91 Building Deformation Monitoring A.2.l.31B March 15, 2020 parameters to be monitored based on the results of the CEB Root Cause, Structural Evaluation and walk downs. Additional parameters monitored will include: alignment of ducting, conduit, and piping; seal integrity; laser target measurements; key seismic gap measurements; and additional instrumentation.

Develop a design standard to implement Aging Management Program B.2.1.3 IB Building Deformation, Program Element 3 - Parameters Monitored/Inspected. The design standard will clarify the deformation evaluation process and provide an auditable format to assess it. The design standard will include steps for each of the three evaluation stages that include parameters monitored, basis for why the parameter is monitored, and conditions that prompts action for the subsequent step.