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10 CFR 50.59 Report, Revision 17 to Updated Final Safety Analysis Report, Revision 14 to Appendix R, Revision 15 to Appendix a, and Revision 146 to the Technical Requirements Manual
	ML16147A196
Person / Time
Site:	Seabrook
Issue date:	04/29/2016
From:	Dean Curtland; NextEra Energy Seabrook
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
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United States Nuclear Regulatory Commission Attention:

Document Control Desk Washington, DC 20555-0001 Seabrook Station NEXTera ENERGY& SEABROOK April 29, 2016 10 CPR 50.59( d)(2) 10 CPR 50.71(e) Docket 50-443 SBK-L-16048 10 CPR 50.59 Report, Revision 17 to the Seabrook Station Updated Final Safety Analysis Report, Revision 14 to Appendix R, "Fire Protection Safe Shutdown Capability," Revisiqn 15 to Appendix A, "Evaluation and Comparison to BTP APCSB 9.5-1," and Revision 146 to the Technical Requirements Manual NextEra Energy Seabrook, LLC (NextEra) encloses the 10 CFR50.59 Report, Revision 17 to the Seabrook Station Updated Final Safety Analysis Report (UFSAR), Revision 14 to Appendix R, "Fire Protection Safe Shutdown Capability," Revision 15 to Appendix A, "Evaluation and Comparison to BTP APCSB 9.5-1," and Revision 146 to the Technical Requirements Manual. The 10 CPR 50.59 Report and the UFSAR are submitted pursuant to the requirements of 10 CPR 50.59(d)(2) and 10 CPR 50.71(e).

The 10 CPR 50.59 report covers the period from August 16, 2014 through February 15, 2016. UFSAR Revision 17 incorporates approved and implemented design changes and UFSAR changes identified through February 15, 2016. The incorporated changes to the UFSAR have been reviewed in accordance with 10 CPR 50.59. The reviews determined that these changes did not require NRC approval.

The UFSAR is provided in its entirety on CD-ROM in Portable Document Format (PDF). Changes from Revision 16 are indicated by a change in number and a vertical line (revision bar) in the margin next to the change. The List of Effective Pages contained within the UFSAR provides a listing of each page and its revision number with a revision bar indicating which pages contain changes. The controlled drawings referenced in the UFSAR are not being provided as they have the potential to contain security-related information.

The drawings are available on site for NRC review. Appendix R, Appendix A and the Technical Requirements Manual are also provided on CD-ROM. NextEra Energy Seabrook, LLC, P.O. Box 300, Lafayette Road, Seabrook, NH 03874 U.S. Nuclear Regulatory Commission SBK-L-16048/Page 2 The summaries of 10 CFR 50.59 evaluations for design changes incorporated in Revision 17 of the UFSAR are attached as Enclosure

1. No 10 CFR 50.59 evaluations were performed for other activities during the reporting period. Enclosure 2 provides a summary for changes to the UFSAR incorporated using the guidance ofNEI 98-03, "Guidelines for Updating Final Safety Analysis Reports." Enclosure 3 is a listing of UFSAR Change Requests (UFCRs) incorporated in UFSAR Revision 17 during the reporting period. The affected Sections, Tables and Figures are provided for each UFCR. Enclosure 4 contains UFCRs incorporated in Appendix R, Enclosure 5 provides the UFCRs incorporated in Appendix A and Enclosure 6 contains UFCRs incorporated in the Technical Requirements Manual. One copy of the UFSAR revision on CD-ROM is being submitted to the Document Control Desk, Washington, DC, along with a copy to the Region I Regional Office and a copy to the Resident Inspector at Seabrook Station. Should you have any questions regarding this matter, please contact Mr. Michael Ossing, Licensing Manager, at (603) 773-7512.

I declare under penalty of perjury that the foregoing is true and correct. Executed on APJt.tl-Z.' '2016 Sincerely, Dean Curtland Site Vice President NextEra Energy Seabrook, LLC cc: NRC Region I Administrator NRC Project Manager, Project Directorate I-2 NRC Senior Resident Inspector Enclosed CD Listing: CD 1, 39 files, 580,001,792 bytes CD 2, 3 files, 68,456,448 bytes Seabrook Station 1 O CFR 50.59 Report, Enclosure 1 Enclosure 1 to SBK-L-16048 Summary Report of Facility Changes, Tests, and Experiments Completed in Accordance with the Requirements of 10CFR50.59 for Revision 17 of the Updated Final Safety Analysis Report Seabrook Station 10 CFR 50.59 Report, Enclosure 1 Design Change Evaluations Design changes documented in the following Engineering Changes (EC) were installed during the period covered by the 10 CFR 50.59 Report. A 10 CFR 50.59 evaluation was performed for the ECs identified below. For each of the evaluations performed, there were no activities requiring prior NRC approval identified.

10CFR50.59 Evaluation 13-002,13-003, and 13-004 EC 279768, Update Design Basis Document for Westinghouse NSAL 07-10, 07-11, 09-01, and 09 02 Summary Description and Purpose: Westinghouse has issued several Nuclear Safety Advisory Letters (NSALs) which document potential safety issues pertaining to Westinghouse accident analyses.

EC 279768 served as the design change vehicle to incorporate the results of these analysis into the Seabrook design basis for the following NSALs: -NSAL-07-10, "Loss-of-Normal Feedwater/Loss-of-Offsite AC Power Analysis PORV Modeling Assumptions," dated 11/07 /07. -NSAL-07-11, "Decay Heat Assumption in Steam Generator Tube Rupture Overfill Analysis Methodology," dated 11/15/07. -NSAL-09-1, "Rod Withdrawal at Power Analysis for Reactor Coolant System Overpressure," dated 2/4/09. -NSAL-09-2, "Locked Rotor Analysis for Reactor Coolant System Overpressure," dated 517109. Note: EC 279768 determined that no changes were required as a result ofNS.f\L-07-11; therefore there is no UFSAR change or 50.50 Evaluation listed. 10CFR50.59 Evaluation 13-002 10CFR50.59 Evaluation 13-002 incorporates NSAL 07-10. The results of the revised LOAC AOR show an adverse impact with respect to the results on the predicted pressurizer water volume with respect to the results of the LOAC AOR. Therefore, the current LOAC AOR does not remain bounding with respect to the acceptance criteria for pressurizer overfilling due to the change in the assumption concerning PORV operation.

The impact, on safety, of this adverse effect on pressurizer water volume is the subject ofthis evaluation.

lOCFR 50.59 Evaluation 13-003 10CFR50.59 Evaluation 13-003 incorporates NSAL 09-01. The results of the revised RW AP AOR show an adverse impact with respect to the results on the predicted RCS pressure response.

Therefore, the current R W AP AOR does not remain bounding with respect to the acceptance criteria for the maximum RCS pressure case. The impact, on safety, of the adverse effect on RCS pressure is the subject ofthis evaluation.

lOCFR 50.59 Evaluation 13-004 1 OCFR50.59 Evaluation 13-004 incorporates NSAL 09-02. 1 of4 Seabrook Station 1 O CFR 50.59 Report, Enclosure 1 The results of the revised LR AOR show an adverse impact with respect to the results on the predicted RCS pressure compared to the previous value. The impact of this adverse effect on RCS pressure is the subject ofthis evaluation.

Evaluation Summary: 1 OCFR50.59 Evaluation 13-002 The evaluation identified one consideration that had the potential to be adverse. The new analysis for LOAC event shows an adverse impact on pressurizer overfill when the PORV s are assumed fail to open. Therefore, the current LOAC AOR, with PORVs assumed open, is no longer bounding.

The evaluation discusses that even with the more adverse results for the LOAC, the safety analysis pressurizer overfill criterion continues to be met with a margin of more than 300 ft3 to pressurizer overfill.

The evaluation concluded that pressurizer overfill is not predicted as a result of the input assumption changes, the other analysis acceptance criteria continues to be met and that prior NRC approval was not required for this design change. 10CFR50.59 Evaluation 13-003 No explicit Seabrook specific analysis existed prior to running the Seabrook specific cases and the result of the sensitivity concluded that 65% reactor power resulted in the highest RCS pressure result (LOL/TT).

These results remain below the acceptance criteria.

Therefore, since the results remain below the acceptance criteria, and no overpressure condition is predicted, the subject activity does not result in a more than minimal increase in the radiological consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR. The evaluation also concluded that the new analysis, using LOFTRAN to analyze RW AP for overpressure is generically approved based on NRC's topic report evaluation and the method is applicable to Westinghouse plants for RW AP analysis.

Also, this analysis is within the scope of the current Seabrook UFSAR methodologies which indicate a range of power levels should be considered.

This evaluation concluded that the results of the revised RW AP AOR were acceptable as the acceptance criteria continues to be met and the code utilized does not result in a departure from a method of evaluation previously approved by the NRC. Prior NRC approval was not required for this design change. 10CFR50.59 Evaluation 13-004 The proposed activity is analytical adding a 45 psi penalty to the Locked Rotor AOR and increases LOL/TT to 2697 .2 psia for predicted RCS peak pressures.

The new LR predicted pressure with the penalty is 2622.5 psia. Both of these new RCS peak pressures remain below the RCS Westinghouse Safety Analysis Limit of2748.5 psia and Technical Specification

2.1.2 value

of2735 psig. The evaluation concluded that there are no restrictions from the subject activity and that prior NRC approval was not required for this design change. 2 of4 Seabrook Station 1 O CFR 50.59 Report, Enclosure 1 10CFR50.59 Evaluation 15-004 EC 283086, Seabrook Cycle 18 Reload Design Summary Description and Purpose: EC 283086 provides the safety evaluation, instructions, and data necessary to refuel and start the Seabrook Cycle 18 core reload. The consecutively long operating windows and coastdown in Cycle 17 and the proposed Cycle 18 design result in the Cycle 18 design exceeding the input assumptions (burnup limits) used in the current fuel handling accident analysis.

This reload implements a reanalysis of the radiological consequences of the fuel handling accident presented in Section 15.7.4.3 of the SBK UFSAR using the Alternative Source Term methodology described Reg Guide 1.183. The change results in an increase in the dose calculation for the fuel handling event, requiring a 5059 evaluation.

Evaluation Summary: A 10 CFR 50.59 evaluation was performed for this modification and concluded that the FHA dose consequence analysis is less than a minimal increase for the EAB, LPZ, and CR envelope.

Additionally, the evaluation concluded that it is acceptable to implement the change in methodology used for the gap release fraction without a license amendment because the new method has already been approved by the NRC for the intended application at another facility.

In accordance with 10 CFR 50.59, it has determined that operation of the Seabrook Cycle 18 reload core can be implemented with no changes to the Technical Specifications and the licensing basis. Therefore, prior NRC approval is not required for the implementation of this design change. 3 of4 Seabrook Station 10 CFR 50.59 Report, Enclosure 1 10CFR50.59 Evaluation 14-001 EC 282474, RCP Shield Shutdown Seal Summary Description and Purpose: Engineering Change 282474 replaced the Reactor Coolant Pump (RCP) number 1 seal insert with a modified design called the SHIELD Shutdown Seal (SDS). The Westinghouse SHIELD Shutdown Seal used is the 3rd Generation Shutdown Seal designed by Westinghouse and is a low leakage RCP seal qualified for the service conditions for 7 days to allow negligible (1 gpm per pump) Reactor Coolant System (RCS) inventory losses through the seals upon a loss of all seal cooling. The SDS is passively activated by a loss of all seal cooling (LOASC) that would occur as a result of a station blackout (SBO) or extended loss of alternating current (ac) power (ELAP) and, therefore, requires no operator action for deployment.

Evaluation Summary: A 10 CFR 50.59 evaluation was performed for this modification.

The evaluation determined that with one exception, all potentially impacted UFSAR described system, structure, or component (SSC) design functions were screened-out.

The RCP design function, to provide core cooling flow during normal operating conditions, was screened-in for further evaluation given it could be adversely impacted ifthe SDS were to inadvertently actuate and that actuation causes an alarm which leads the operators to shutdown the pump. The UFSAR Sections 5.4.1.3.d, 15.3.3 and 15.3.4 describe an RCP shaft seizure (i.e., locked rotor) or the break of a RCP shaft which is an ANS Condition IV limiting fault. Seizure of the shaft/rotor would cause the flow through the affected reactor coolant loop to be rapidly reduced, leading to an initiation of a reactor trip on a low flow signal. An inadvertent SDS actuation would provide a negligible amount of resistance and wear against the rotating shaft that is incapable of seizing or breaking the RCP shaft. If the SDS were to actuate inadvertently, it would not increase the probability of a RCP shaft seizure or break and, therefore, would not change the frequency of occurrence of this postulated accident.

Even in the unlikely case that an SDS actuation causes an alarm which leads the operators to shutdown the pump due to a debris-related seal leakoff alarm the malfunction result (i.e., its impact on core cooling) is no different than previously evaluated in the UFSAR (i.e., Complete Loss of Forced Reactor Coolant Flow, 15.3.2, and Partial Loss of Forced Reactor Coolant Flow, 15.3.1). The evaluation concluded that the modification did not require prior NRC approval.

4 of4 Seabrook Station 1 O CFR 50.59 Report, Enclosure 2 Enclosure 2 to SBK-L-16048 Summary of Changes to the Updated Final Safety Analysis Report Incorporated Using the Guidance ofNEI 98-03, "Guidelines for Updating Final Safety Analysis Reports" Seabrook Station 10 CFR 50.59 Report, Enclosure 2 The following provides a summary of changes incorporated in Revision 17 of the Updated Final Safety Analysis Report using the guidance contained in NEI 98-03, "Guidelines for Updating Final Safety Analysis Reports." The summaries provide the UFSAR Change Request (UFCR/UCR) number, affected Sections, Tables, or Figures and a description of the change. UFCR/UCR Number EC281849 Affected Sections, Tables, and Figures 9.1.3.5 Description of Change: This change installs reliable primary and backup level instrumentation to monitor Spent Fuel Pool (SFP) water level in accordance with the requirements ofNRC issued Order EA-12-051, Issuance of Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, March 12, 2012. A new UFSAR subsection will be added to describe the beyond design basis design :function of the new Spent Fuel Pool Instrumentation system. The addition ofthis description is considered to be editorial since the SFPIS is considered to be beyond design basis. EC283851 Figure 2.1-5 and 2.1-4 Description of Change: This change updated Figure 2.1-5 (Station Layout) and Figure 2.1-4 (Site Boundaries).

The removal of the buildings from the UFSAR figures, updates the UFSAR to reflect the current site configuration.

The correction to the figures to reflect current site structures does not change the meaning or substance of the information in the UFSAR. EC285398 12.5.1 Description of Change: AR222295/UFCR10-007 revised the UFSAR to reference ANSI/ANS 3.1-1993 Qualification and Training of Personnel for Nuclear Power Plants (with two exceptions as noted in UFSAR Section 1.8). UFSAR Chapter 12.5.l still incorrectly references ANSI/ANS 3.1-1981 and also ANSI/ANS 3.1-1978.

This change eliminates the discrepancy between UFSAR section 13.1 and 12.5 by removing the references to the 1978 and 1981 ANSI/ ANS standards and referring the reader to section 13 .1.3 .1 for personnel qualification requirements.

This change aligns with item 4.2.2.F(l)c in EN-AA-203-1201, "Correction of inconsistencies within the UFSAR (e.g., between sections)," and is excluded from review under 10 CFR 50.59. 1of2 Seabrook Station 10 CFR 50.59 Report, Enclosure 2 UFCR/UCR Number EC285392 Affected Sections, Tables, and Figures Figure 7.6-6 Description of Change: Review ofUFSAR Figure 7.6-6, "Diagram Showing Generating Plant Processing for Low Temperature Interlocks for RCS Pressure Control," determined that tag ID's were not shown correctly for the isolation devices and the Train B Protection set. This UCR provides editorial corrections to the UFSAR Figure 7.6-6 based on existing drawings.

2 of2 Seabrook Station 10 CFR 50.59 Report, Enclosure 3 Enclosure 3 to SBK-L-16048 Listing ofUFSAR Change Requests (UFCRs/UCRs)

Incorporated in Updated Final Safety Analysis Report, Revision 17 Seabrook Station 10 CFR 50.59 Report, Enclosure 3 UFCR/UCR Affected Sections Affected Tables Affected Figures Number 09-019 2.1-5 EC2217 2.3-36 EC276930 1.2.3.4, 4.4.6.1, 7.7.1.9 4.4-7, 7.7-9 EC279387 8.2.1.4.c, 8.2.1.4.g EC279768 7.2, 15.2, 15.4 7.2-4, 15.2-1, 15.3-1, 15.2-1, 15.2-5 15.3-2 EC280335 5.2.3.2.a, 9.3.4.1.d, 9.3.4.2.b, 9 .3 .4.2.f.2.a EC280504 8.3.2.1 EC280653 2.3.3.3, Appendix 7A 2.3-37. 7.5-1 EC281842 5.4.14.2.b.3 EC281849 9.1.3.5 EC282474 5.4, EC282641 9.1.2.1, 9.1.2.3, 9.1.6 9.1-22 EC283086 1.2, 1.8, 4.2, 4.3, 15.0, 15.3, 15.6, 15.7, Appendix 15C EC283092 1.8 EC283158 9.3 EC283287 4.3, 7.7 1-6.1 1of2 Seabrook Station 10 CFR 50.59 Report, Enclosure 3 UFCR/UCR Number EC283325 EC283851 EC283978 EC283981 EC284008 EC284435 EC284635 EC284737 EC284767 EC285353 EC285392 EC285398 EC285415 Affected Sections Affected Tables 2.2 2.2-1 1.8, 8.1, 8.3, 10.4 3.9(B)-27 9.1.1.1, 9.1.1.2, 9.1.1.3, 9.1.6 10.4.1.5 7.7.1.1 9.1.5.1 8.3.1.1.g 2.4, 3.4 12.5.1 13.2.1.2 2 of2 Affected Figures 2.2-1, 2.2-5 2.1-5, 2.1-4 9.1-13, 9.1-14, 9.1-15, 9.1-16, 9.1-17, 9.1-18 7.6-6 Seabrook Station 1 OCFRS0.59 Report, Enclosure 4 Enclosure 4 to SBK-L-16048 Listing of UFSAR Change Requests (UFCRs) Incorporated in Appendix R, Fire Protection Safe Shutdown Capability, Revision 14 Seabrook Station 1 OCFRS0.59 Report, Enclosure 4 UFCR/UCR Affected Sections Affected Tables/ Affected Figures Number Tabulations EC276930 MCR 3.1.3.4-5, MCR 3.1.3.4-6, MCR3.l.3.4-7 EC281148 RSS 3.1.3.4 EC281882 MCR3.l.3.4 EC285356 3.3.3.4 1of1 Seabrook Station 1 OCFRS0.59 Report, Enclosure 5 Enclosure 5 to SBK-L-16048 Listing ofUFSAR Change Requests (UFCRs) Incorporated in Appendix A, Evaluation and Comparison to BTP APCSB 9.5-1, Revision 15 Seabrook Station 1 OCFR50.59 Report, Enclosure 5 UFCR/UCR Affected Sections Affected Tables Affected Figures Number EC280451 F.2, Tab 11 EC281647 F.2, Tab 7; F.2, Tab 8; F.2, Tab 13 EC281896 F.2, Tab 11 EC282368 F.2, Tab 11 EC282581 F.2, Tab 7 EC283274 F.2, Tab 11 EC284930 F.2, Tab 7 EC285591 F.3 1of1 Seabrook Station 1 OCFR50.59 Report, Enclosure 6 Enclosure 6 to SBK-L-16048 Listing ofUFSAR Change Requests (UFCRs) Incorporated in Revisions 143 through 146 of the Technical Requirements Manual

.. Seabrook Station 1 OCFRS0.59 Report, Enclosure 6 UFCR/UCR Number EC276930 EC283086 EC283287 EC283642 Affected Sections TR 20-3.3.3.2 Chapter 6, COLR TR 20; Chapter 6, COLR TR22 1of1 Affected Tables Affected Figures

United States Nuclear Regulatory Commission Attention: