Text

Issuance of Amendments Regarding Implementation of the Protected Service Water System
	ML14206A790
Person / Time
Site:	Oconee
Issue date:	08/13/2014
From:	Hall J; Plant Licensing Branch II
To:	Batson S; Duke Energy Carolinas
	Hall J
References
	TAC ME7737, TAC ME7738, TAC ME7739, TAC ME7746, TAC ME7747, TAC ME7748
	Download: ML14206A790 (86)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Scott Batson Site Vice President Oconee Nuclear Station Duke Energy Carolinas, LLC 7800 Rochester Highway Seneca, SC 29672-0752 August 13, 2014

SUBJECT:

, OCONEE NUCLEAR STATION, UNITS 1, 2 AND 3, ISSUANCE OF AMENDMENTS REGARDING IMPLEMENTATION OF THE PROTECTED SERVICE WATER SYSTEM (TAC NOS. ME7737, ME7738, ME7739, ME7746, ME7747, AND ME7748)

Dear Mr. Batson:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment Nos. 386,.

388, and 387 to Renewed Facility Operating Licenses DPR-38, DPR-47, and DPR-55, for the Oconee Nuclear Station (ONS), Units 1, 2, and 3, respectively. The amendments consist of changes to the ONS operating licenses, Technical Specifications (TSs), and Updated Final Safety Analysis Report (UFSAR) in response to the application from Duke Energy Carolinas, LLC (Duke Energy), dated December 16, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML120030226), as supplemented by letters dated January 20, 1 March 1,2 March 16,3 April 18,4 July 11,5 July 20, 6 August 31, 7 and November 2, 2012, 8; April 5, 9 June 28,10 August 7, 11 and December 18, 2013;12 and February 14,13 April 3, 14 April 11 15, and July 24, 2014. 16 These amendments add a License Condition and revise the TSs and UFSAR for ONS, Units 1, 2, and 3, to add the new Protected Service Water (PSW) system to the plant's licensing basis as an additional method of achieving and maintaining safe shutdown of the reactors in the event of a high-energy line break or a fire in the Turbine Building, which is shared by afl three units.

A copy of the NRC staff's related Safety Evaluation (SE) providing the technical bases for the staff's approval of the amendments is also enclosed. These amendments and the related SE do not approve nor endorse the as-in~talled PSW electrical system cable configurations. Duke 1 ADAMS Accession No. ML12025A124 2 ADAMS Accession No. ML12080A199 3 ADAMS Accession No. ML12081A126 4 ADAMS Accession No. ML12110A175 5 ADAMS Accession No. ML12195A325 6 ADAMS Accession No. ML12207A109 7 ADAMS Accession No. ML12249A400 8 ADAMS Accession No. ML12312A031 9ADAMSAccession No. ML13123A159 10 ADAMS Accession No. ML13190A016 11 ADAMS Accession No. ML13228A268 12 ADAMS Accession No. ML13358A042 13 ADAMS Accession No. ML14055A068 14 ADAMS Accession No. ML14099A264 15 ADAMS Accession No. ML14107A034 16 ADAMS Accession No. ML14210A356

Energy analyzed the PSW electrical system* configurations and installed the PSW electrical cables and power supplies under the provisions of 10 CFR 50.59, and thus those parts of the system were not included in the ~cope of the NRC staff's review for these amendments. The installed configurations of the PSW cabling and associated on site power supply systems are the subject of a* pending NRC inspection activity, as documented as an unresolved item in the NRC Component Design Basis Inspection Report, dated June 27, 2014, Section 1.2.b.v. (ADAMS Accession No. ML14178A535).

A Notice of Issuance will be included in the Commission's biweekly Federal Register notice. If you have any questions, please contact me at 301-415-4032.

Docket Nos. 50-269, 50-270, and 50-287

Enclosures:

1. Amendment No. 386 to DPR-38

2. Amendment No. 388 to DPR-47

3. Amendment No. 387 to DPR-55

4. Safety Evaluation cc w/encls: Distribution via Listserv Sincerely, es R. Hall, Senior Project Manager Pant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

UNITED STATES

.NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-269 OCONEE NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 386 Renewed License No. DPR-38

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to the Oconee Nuclear Station, Unit 1 (the facility),

Renewed Facility Operating License No. DPR-38, filed by Duke Energy Carolinas, LLC (the licensee), dated December 16, 2011, as supplemented by letters dated January 20, March 1, March 16, April 18, July 11, July 20, August 31, and November 2, 2012, Aprii.J5, June 28, August 7, and December 18, 2013, and February 14, April 3, April11, and July 24, 2014, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.8 of Renewed Facility Operating License No. DPR-38 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 386, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

Accordingly, the license is hereby amended by changing the Renewed Facility Operating License No. DPR-38 License Condition 3.1 to read as follows:

1 I.

Protected Service Water System Seismic Assessment License Condition Duke Energy Carolinas, LLC (Duke Energy) shall perform a seismic probabilistic risk assessment (SPRA) which includes the Protected Service Water (PSW) system, in accordance with the Electric Power Research Institute (EPRI) Report No. 1025287, "Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1:

Seismic," (i.e., the SPID report, November 2012) for the Oconee Nuclear Station (ONS). Duke Energy shall expand the Seismic Equipment List (SEL) to include the PSW system.

4.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.

Attachment:

Changes to Renewed Facility Operating License No. DPR-38 and the Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION

. Lantz, 'ctin for Regions II & Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance:.,A.ugus t 13, 2014

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 388 Renewed License No. DPR-47

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to the Oconee Nuclear Station, Unit 2 (the facility),

Renewed Facility Operating License No. DPR-47, filed by Duke Energy Carolinas, LLC (the licensee), dated December 16, 2011, as supplemented by letters dated January 20, March 1, March 16, April 18, July 11, July 20, August 31, and November 2, 2012, April 5, June 28, August 7, and December 18, 2013, and February 14, April 3, April 11, and July 24, 2014, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2..

Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.8 of Renewed Facility Operating License No. DPR-47 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 388, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

Accordingly, the license is hereby amended by changing the Renewed Facility Operating License No. DPR-47 License Condition 3.1 to read as follows:

1 I.

Protected Service Water System Seismic Assessment License Condition Duke Energy Carolinas, LLC (Duke Energy) shall perform a seismic probabilistic risk assessment (SPRA) which includes the Protected Ser-Vice Water (PSW) system, in accordance with the Electric Power Research Institute (EPRI) Report No. 1025287, "Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1:

Seismic," (i.e., the SPID report, November 2012) for the Oconee Nuclear Station (ONS). Duke Energy shall expand the Seismic Equipment List (SEL) to include the PSW system.

4.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.

Attachment:

Changes to Renewed Facility Operating License No. DPR-47 and the Technical Specifications

. FOR THE NUCLEAR REGULATORY COMMISSION Ryan E. Lantz, A inu._~I!'IL for Regions II & Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: August 13, 2014

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 387 Renewed License No. DPR-55

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to the Oconee Nuclear Station, Unit 3 (the facility),

Renewed Facility Operating License No. DPR-55, filed by Duke Energy Carolinas, LLC (the licensee), dated December 16, 2011, as supplemented by letters dated January 20, March 1, March 16, April18, July 11, July 20, August 31, and November 2, 2012, April 5, June 28, August 7, and December 18, 2013, and February 14, April 3, April 11, and July 24, 2014, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended {the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.8 of Renewed Facility Operating License No. DPR-55 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 387, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

Accordingly, the license is hereby amended by changing the Renewed Facility Operating 1

License No. DPR-55 License Condition 3.1 to read as follows:

I.

Protected Service Water System Seismic Assessment License Condition Duke Energy Carolinas, LLC (Duke Energy) shall perform a seismic probabilistic risk assessment (SPRA) which includes the Protected Service Water (PSW) system, in accordance with the Electric Power Research Institute (EPRI) Report No. 1025287, "Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1:

Seismic," (i.e., the SPID report, November 2012) for the Oconee Nuclear Station (ONS). Duke Energy shall expand the Seismic Equipment List (SEL) to include the PSW system.

4.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.

Attachment:

Changes to Renewed Facility Operating License No. DPR-55 and the Technical Specifications FOR THE NUCLEAR REGULA TORY COMMISSION

. Lantz, A ting'1""Rl=---mr for Regions II & Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: August 13, 2014

ATTACHMENT TO LICENSE AMENDMENT NO. 386 RENEWED FACILITY OPERATING LICENSE NO. DPR-38 DOCKET NO. 50-269 LICENSE AMENDMENT NO. 388 RENEWED FACILITY OPERATING LICENSE NO. DPR-47 DOCKET NO. 50-270

AND LICENSE AMENDMENT NO. 387 RENEWED FACILITY OPERATING LICENSE NO. DPR-55 DOCKET NO. 50-287 Replace the following pages of the Renewed Eacility Operating Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Licenses License No. DPR-38, page 3 License No. DPR-38, page 11 License No. DPR-47, page 3 License No. DPR-47, page 11 License ~o. DPR-55, page 3 License No. DPR-55, page 11 TSs Table of Contents Page iii Table of Contents Page iv Table of Contents Page v Page 3.7.10-1 Page 5.0-23 Page 5.0-24 Page 5.0-25 Page 5.0-26 Page 5.0-27 Insert Pages Licenses License No. DPR-38, page 3 License No. DPR-38, page 11 License No. DPR-47, page 3 License No. DPR-47,* page 11 License No. DPR-55, page 3 License No. DPR-55, page 11 TSs Table of Contents Page iii Table of Contents Page iv Table of Contents Page v Page 3.7.10-1 Pages 3.7.10-2 through 3.7.10-3 Pages 3.7.10a-1 through 3.7.10a-4 Page 5.0-23 Page 5.0-24 Page 5.0-25 Page 5.0-26 Page 5.0-27 Page 5.0-28 A.

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal.

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 386 are hereby incorporated in the license. The licensee shall operate the facility in accordance with*the Technical Specifications.

C.

This license is subject to the following antitrust conditions:

Applicant makes the commitment~ contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production.~nd sale of electricity.

Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in 1f1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.

1.

As used herein:

(a)

"Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub.::transmission voltage by one electric system to another.

(b)

"Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of Renewed License No. DPR-38 Amendment No. 386 (b)

Operations to mitigate fuel damage considering the following:

1.

Protection and use of personnel assets

2.

Communications

3.

Minimizing fire spread

4.

Procedures for implementing integrated fire response strategy

5.

Identification of readily-available pre-staged equipment

6.

Training on integrated fire response strategy

7.

SFP mitigation measures (c)

Actions to minimize release to include consideration of:

1.

Water spray scrubbing

2.

Dose to onsite responders I.

Protected Service Water System Seismic Assessment License Condition Duke Energy.Carolinas, LLC (Duke Energy) shall perform a seismic probabilistic ri$k assessment (SPRA) which includes the Protected Service Water (PSW) system, in accordance with the Electric Power Research Institute (EPRI) Report No. 1025287, "Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic," (i.e., the SPID report, November 2012) for the Oconee Nuclear Station (ONS). Duke Energy shall expand the Seismic Equipment List (SEL) to include the PSW system.

4.

This renewed license is effective as of the date of issuance and shall expire at midnight on February 6, 2033.

Attachment:

FOR THE NUCLEAR REGULA TORY COMMISSION Original signed by Roy P. Zimmerman Roy Zimmerman, Acting Director Office of Nuclear Reactor Regulation

1) Appendix A-Technical Specifications Renewed License No. DPR-38 Date of Issuance: May 23, 2000 Renewed License No. DPR-38 Amendment No. 386 A.

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal.

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 388 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

C.

This license is subject to the following antitrust conditions:

Applicant makes ~he commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.

Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described wtiich, on balance; provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ~1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.

1.

As used herein:

(a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.

(b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of Renewed License No. DPR-47 Amendment No. 388 (b)

Operations to mitigate fuel damage considering the following:

1. Protection and use of personnel assets

2. Communications

3. Minimizing fire spread

4. Procedures for implementing integrated fire response strategy

5. Identification of readily-available pre-staged equipment

6. Training on integrated fire response strategy

7. SFP mitigation measures (c)

Actions to minimize release to include consideration of:

1. Water spray scrubbing

2. Dose to onsite responders I.

Protected Service Water System Seismic Assessment License Condition Duke Enerby Carolinas, LLC (Duke Energy) shall perform a seism 1

ic probabilistic risk assessment (SPRA) which includes the Protected Service Water (PSW) system, in accordance with the Electric Power Research Institute (EPRI) Report No. 1025287, "Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic," (i.e., the SPID report, November 2012) for the Oconee Nuclear Station (ONS). Duke Energy shall expand the Seismic Equipment List (SEL) to include the PSW system.

1

4.

This renewed license is effective as of the date of issuance and shall expire at midnight on October 6, 2033.

Attachment:

FOR THE NUCLEAR REGULATORY COMMISSION Original signed by Roy P. Zimmerman Roy Zimmerman, Acting Director Office of Nuclear Reactor Regulation

\\

1) Appendix A-Technical Specifications Renewed License No. DPR-47 Date of Issuance: May 23, 2000 Renewed License No. DPR-47 Amendment No. 388 A.

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal.

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 38T are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

C.

This license is subject to the following antitrust conditions:

Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.

Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ~1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.

1.

As used herein:

(a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.

(b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a

-lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of Renewed License No. DPR-55 Amendment No. 387 (b)

Operations to mitigate fuel damage considering the following:

1. Protection and use of personnel assets

2. Communications

3. Minimizing fire spread

4. Procedures for implementing integrated fire response strategy

5. Identification of readily-available pre-staged equipment

6. Training on integrated fire response strategy

7. SFP mitigation measures (c)

Actions to minimize release to include consideration of:

1. Water spray scrubbing

2. Dose to onsite responders I.

Protected Service Water System Seismic Assessment License Condition Duke Energy Carolinas, LLC (Duke Energy) shalll perform a sei.smic probabilistic risk assessment (SPRA) which includes the Protected Service Water (PSW) system, in accordance with the Electric Power Research Institute (EPRI) Report No. 1025287, "Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic," (i.e., the SPID report, November 2012) for the Oconee Nuclear Station (ONS). Duke Energy shall expand the Seismic Equipment List (SEL) to include the PSW system.

4.

This renewed license is effective as of the date of issuance and shall expire at midnight on July 19, 2034.

Attachment:

FOR THE NUCLEAR REGULATORY COMMISSION Original signed by Roy P. Zimmerman Roy Zimmerman, Acting Director Office of Nuclear Reactor Regulation

1) Appendix A-Technical Specifications Renewed License No. DPR-55 Date of Issuance: May 23, 2000 Renewed License No. DPR-55 Amendment No. 387

TABLE OF CONTENTS 3.4.6 3.4.7 3.4.8 3.4.9 3.4.10 3.4.11 3.4.12 3.4.13 3.4.14 3.4.15 3.4.16 3.5 3.5.1 3.5.2 3.5.3 3.5.4 3.6 3.6.1 3.6.2 3.6.3 3.6.4 3.6.5 3.7 3.7.1 3.7.2 3.7.3 3.7.4 3.7.5 3.7.6 3.7.7 3.7.8 3.7.9 3.7..10 3.7.10a 3.7.11 3.7.12 3.7.13 RCS Loops - MODE 4........................................... ;.......................... 3.4.6-1 RCS Loops - MODE 5, Loops Filled....... :......................................... 3.4.7-1 RCS Loops -

MODE 5, Loops Not Filled........................................... 3.4.8-1 Pressurizer.......................................................................................... 3.4. 9-1 Pressurizer Safety Valves................................................................... 3.4.10-1 RCS Specific Activity.......................................................................... 3.4.11-1 Low Temperature Overpressure Protection (LTOP)

System.......................................................................................... 3.4.12-1 RCS Operational LEAKAGE.,............................................ :................ 3.4.13-1 RCS Pressure Isolation Valve (PIV) Leakage..................................... 3.4.14-1 RCS Leakage Detection Instrumentation.......................*.................... 3.4.15-1 Steam Generator (SG) Tube lntegrity................................................. 3.4.16-1 EMERGENCY CORE COOLING SYSTEMS (ECCS)................................ 3.5.1-1,

Core Flood Tanks (CFTs)....... :........................................................... 3.5.1-1

('

High Pressure lnjection.............................................. ~........................ 3.5.2-1 Low Pressure lnjection............................................... ~........................ 3.5.3-1 Borated Water Storage Tank (BWST)................................................ 3.5.4-1 CONTAINMENT SYSTEMS....................................................................... 3.6.1-1 Containment....................................................................................... 3.6.1-1 Containment Air Locks........................................................................ 3.6.2-1 Containment Isolation Valves............................................................. 3.6.3-1 Containment Pressure *"******************************....................................... 3.6.4-1 Reactor Building Spray and Cooling System...................................... 3.6.5-1 PLANT SYSTEMS..........-................... *......................................................... 3.7.1-1 Main Steam Relief Valves (MSRVs)................................................... 3.7.1-1 Turbine Stop Valves (TSVs)............................................................... 3.7.2-1 Main Feedwater Control Valves (MFCVs ), and Startup Feedwater Control Valves (SFCVs).............................................. 3.7.3-1 Atmospheric Dump Valve (ADV) Flow Paths...................................... 3.7.4-1 Emergency Feedwater (EFW) System.............................. :**************.. 3.7.5-1 Upper Surge Tank (UST) and Hotwell (HW)....................................... 3.7.6-1 Low Pressure Service Water (LPSW) System.................................... 3.7.7-1 Emergency Condenser Circulating Water (ECCW).......... :................. 3.7.8-1 Control Room Ventilation System (CRVS) Booster Fans................................................................... 3.7.9-1 Protected Service Water (PSW)......................................................... 3.7.10-1 Protected Service Water (PSW) Battery Cell Parameters................................................................................... 3.7.1 Oa-1 Spent Fuel Pool Water Level.............................................................. 3.7.11-1 Spent Fuel Pool Boron Concentration................................................. 3.7.12-1 Fuel Assembly Storage....................................................................... 3.7.13-1 OCONEE UNITS 1, 2, & 3 iii Amendment Nos. 386, 388, & 387

TABLE OF CONTENTS 3.7.14 3.7.15 3.7.16 3.7.17 3.7.18 3.7.19 3.8 3.8.1 3.8.2 3.8.3 3.8.4 3.8.5 3.8.6 3.8.7 3.8.8 3.8.9 3.9 3.9.1 3.9.2 3.9.3 3.9.4 3.9.5 3.9.6 3.9.7 3.9.8 3.10 3.10.1 3.10.2 4.0 4.1 4.2 4.3 4.4 5.0 5.1 Secondary Specific Activity........................................................... 3.7.14-1 Decay Time for Fuel Assemblies in Spent Fuel Pool (SFP)*............................................................................. 3.7.15-1 Control Room Area Cooling Systems (CRACS)........................... 3.7.16-1 Spent Fuel Pool Ventilation System (SFPVS)............................... 3.7.1T-1 Dry Spent Fuel Storage Cask Loading and Unloading........ ;......... 3.7.18-1 Spent Fuel Pool Cooling (SFPC) Purification System Isolation from Borated Water Storage Tank (BWST)............. 3.7.19-1 ELECTRICAL POWER SYSTEMS..................................................... 3.8:1-1 AC Sources-Operating............................................................... 3.8.1-1 AC Sources.;.... Shutdown............................................................... 3.8.2-1 DC Sources-Operating.............................................. ~................ 3.8.3-1 DC Sources-Shutdown.............................................................. 3.8.4-1 Battery Cell Parameters................................................................ 3.8.5-1 Vital Inverters-Operating................ ~........................................... 3.8.6-1 Vital Inverters-Shutdown............................................................ 3.8.7-1 Distribution Systems-Operating.................................................. 3.8.8-1 Distribution Systems-Shutdown................................................. 3.8.9-1 REFUELING OPERATIONS.............................................................. 3.9.1-1 Boron Concentration..................................................................... 3.9.1-1 Nuclear lnstrumentation................................................................. 3.9.2-1 Containment Penetrations............................................................ 3.9.3-1 Decay Heat Removal (DHR) and Coolant Circulation-High Water Level................................................ 3.9.4-1 Decay Heat Removal (DHR) and Coolant Circulation-Low Water Level................................................ 3.9.5-1 Fuel Transfer Canal Water Level.................................................. 3.9.6-1 Unborated Water Source Isolation Valves.................................... 3.9.7-1 Reverse Osmosis (RO) System Operating Restrictions For Spent Fuel Pool (SFP)..................................................... 3.9.8-1 STANDBY SHUTDOWN FACILITY.................................................... 3.10.1-1 Standby Shutdown Facility (SSF)................................................. 3.1 0.1-1 Standby Shutdown Facility (SSF) Battery Cell Parameters............................................................................ 3.10.2-1

DESIGN FEATURES......................................................................... A.0-1 Site Location................................................................................. 4.0-1 Reactor Core................................................................................ 4.0-1 Fuel Storage... :................................................ :............................ 4.0-1 Dry Spent Fuel Storage Cask Loading and Unloading..................4.0-3 ADMINISTRATIVE CONTROLS......................................................... 5.0-1 Responsibility................................................................................. 5.0-1 OCONEE UNITS 1, 2, & 3 iv Amendment Nos. 386, 388, & 387

TABLE OF CONTENTS 5.2 5.3 5.4 5.5 5.6 Organization............................................................................................... 5.0-2 Station Staff Qualifications.......................................................................... 5.0-5 Procedures................................................................................................ 5.0-6 Programs and Manuals.............................................................................. 5.0-7 Reporting Requirements............................................................................. 5.0-24 OCONEE UNITS 1, 2, & 3 v

Amendment Nos. 386, 388, & 387

3. 7 PLANT SYSTEMS 3.7.10 Protected Service Water (PSW) System LCO 3.7.10 The PSW system shall be OPERABLE PSW System 3.7.10

N()TE-----------------------------------------------------------

Not applicable to Unit(s) until startup from a refueling outage after completion of PSW modifications and after all of the PSW system equipment installed has been tested.

APPLICABILITY:

MODES 1 and 2.

ACTIONS i

I

N()TE-----------------------------------------------------------

LC() 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. PSW system is inoperable.

A.1 Restore PSW system to 14 days

. OPERABLE status.

'8. PSW system is inoperable.

B.1 Restore PSW system to

'7 days OPERABLE status.

AND The Standby Shutdown Facility (SSF) is inoperable.

C. ----------------NOTE-----------------

Condition may only be entered I

when contingency measures have been implemented.

Required Action and C.1 Restore PSW system to 30 days from discovery associated Completion Time of OPERABLE status.

of initial inoperability.

Condition A or B not met.

D. Required Action and D.1 Be in M()DE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition C not met.

(continued)

OC()NEE UNITS 1, 2, & 3 3.7.10-1 Amendment Nos. 386, 388, & 387

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR3.7.10.1 Verify the required PSW battery terminal voltage is greater than or equal to the minimur:n established float voltage.

SR 3.7.10.2 Verify the required Keowee Hydroelectric Station power supply can be aligned to and power the PSW electrical system.

SR 3.7.10.3 Verify developed head of PSW primary and booster pumps at flow test point is greater than or equal to the required developed head.

I I

I SR 3.7.10.4 Verify PSW battery capacity of the required battery is adequate to supply, and maintain in OPERABLE status, required emergency loads for the design duty cycle when subjected to a battery service test.

SR 3.7.10.5 Verify the required PSW battery charger supplies

300 amps at greater than or equal to the minimum established float voltage for > 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

OR Verify the required battery charger can recharge the battery to the fully charged state within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while supplying the largest combined demands of the various continuous steady state

loads, after a battery discharge to the bounding PSW event discharge state.

SR 3.7.10.6

N 0 T E -----------------------------

Both HPI pump motor,s are individually tested although only one (1) HPI pump motor is required to support PSW system OPERABILITY.

Verify that the required PSW switchgear and transfer switches can be aligned and power both the "A" and "B" HPI pump motors.

SR 3.7.10.7 Perform functional test of required power transfer switches used for pressurizer heaters, PSW control, electrical panels, vital I&C chargers, and valves.

PSW System 3.7.10 FREQUENCY In accordance with the Surveillance* Frequency Control Program.

In accordance with the Surveillance Frequency Control Program.

In accordance with the lnservice Testing Program.

In accordance with the Surveillance Frequency Control Program.

(continued)

OCONEE UNITS 1, 2, & 3 3.7.10-2 Amendment Nos. 386, 388, & 387

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE SR 3.7.10.8

NOliE-----------------------------

Cooling water flows to the HPI pump motors are individually tested although only flow to the HPI pump motor aligned to PSW power is required to support PSW system OPERABILiliY.

Verify PSW booster pump and valves can provide adequate cooling water flow to HPI pump motor coolers.

SR 3.7.10.9 Verify developed head of PSW portable pump at the flow test point is greater than or equ'al to required developed head.

SR 3.7.10.10 Verify the required PSW valves are tested in accordance with the lnservice liest Program.

SR 3.7.10.11 Perform CHANNEL CHECK for each required PSW instrument channel.

SR 3.7.10.12 Perform CHANNEL CALl BRA liiON for each required PSW instrument channel.

SR 3.7.10.13 Verify for the required PSW battery that the cells, cell plates and racks show no visual indication of physical damage or abnormal deterioration that could degrade battery performance.

PSW System 3.7.10 FREQUENCY In accordance with the lnservice Testing Program.

In accordance with the Surveillance Frequency Control Program.

In accordance with the lnservice liesting Program.

In accordance with the Surveillance Frequency Control Program.

OCONEE UNITS 1, 2, & 3 3.7.1 0-3 Amendment Nos. 386, 388, & 387

3. 7 PLANT SYSTEMS PSW Battery Cell Parameters 3.7.10a 3.7.10a Protected Service Water (PSW) Battery Cell Parameters LCO 3. 7.1 Oa Battery Cell parameters for the required PSW battery shall be within limits.

APPLICABILITY:

When the PSW system is required to be OPERABLE.

ACTIONS

~-----------------------------------------------N()TE----------------------------------~------------------------

LC() 3.0.4 is not applicable.

CONDITI()N REQUIRED ACTI()N COMPLETION TIME A.

Required battery with A.1 Perform SR 3.7.10.1 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months one or more battery cell float voltages :::; 2.07 V.

AND A.2 Perform SR 3.7.10a.1.

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months AND A.3 Restore affected cell 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months voltage > 2.07 V.

B.

Required battery with B.1 Perform SR 3.7.10.1 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months float current > 2 amps.

AND B.2 Restore battery float 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months current to :::; 2 amps.

\\

(continued)

OC()NEE UNITS 1, 2, & 3 3.7.10a-1 Amendment Nos. 386, 388, & 387

ACTIONS (continued)

CONDITION

NOTE------------

Required Actions C.1 and C.2 shall be completed if electrolyte level was below the top of plates.

C.

Required battery with one or more cells electrolyte level less than minimum established design limits.

D.

Required battery with pilot cell electrolyte temperature less than minimum established design limits.

OCONEE UNITS 1, 2, & 3 REQUIRED ACTION

N 0 T E --------------

Required Actions C.1 and C.2 are only applicable if electrolyte level was below the top of plates.

C.1 Restore electrolyte level to above top of plates.

AND]

C.2 Verify no evidence of leakage.

AND C.3 Restore electrolyte level to greater than or equal to minimum established design limits.

D.1 Restore battery pilot cell -

temperature to greater than or equal to minimum established design limits.

PSW Battery Cell Parameters 3.7.10a COMPLETION TIME 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months I

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 31 days 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

( contrnued) 3.7.10a-2 Amendment Nos. 386, 388, & 387

ACTIONS (continued)

CONDITION REQUIRED ACTION E.

Required Action and associated Completion Time of Condition A, B, C, or D not met.

E.1 Declare associated battery inoperable.

OR Required battery with one or more battery cells float voltage

2.07 V and float current > 2 amps.

SURVEILLANCE REQUIREMENTS SR 3.7.10a.1 SR 3.7.10a.2 SR 3.7.10a.3 SR3.7.10a.4 SURVEILLANCE

NOTE--------------------------

Not required to be met when battery terminal voltage is less than the minimum established float voltage of SR 3.7.10.1.

Verify battery float current is ::;; 2 amps.

Verify battery pilot cell voltage is> 2.07 V.

Verify battery connected cell electrolyte level is greater than or equal to minimum established design limits.

Verify battery pilot cell temperature is greater than or equal to minimum established design limits.

PSW Battery Cell Parameters 3.7.10a COMPLETION TIME Immediately FREQUENCY In accordance with the Surveillance Frequency Control Program.

In accordance with the Surveillance Frequency Control Program.

(continued)

OCONEE UNITS 1, 2, & 3 3.7.10a-3 Amendment Nos. 386, 388, & 387

SURVEILLANCE REQUIREMENTS SR 3.7.10a.5 SR 3.7.10a.6 Verify battery connected cell voltage is

> 2.07 V.

Verify battery capacity is ~ 80% of the manufacturer's rating when subjected to a performance discharge test or a modified performance discharge test.

PSW Battery Cell Parameters 3.7.10a In accordance with the Surveillance Frequency Control Program.

In accordance with the Surveillance Frequency Control Program.

AND 12 months when battery shows degradation or has reached 85% of the expect~d life_with capacity < 1 00% of manufacturer's rating.

AND 24 months when battery has reached 85% of the expected life with capacity ~ 1 00% of manufacturer's rating.

OCONEE UNITS 1, 2, & 3 3.7.10a-4,

Amendment Nos. 386, 388, & 387

5.5 Programs and Manuals Programs and Manuals 5.5 5.5.22 Protected Service Water System Battery Monitoring and Maintenance Program This program is applicable only to the Protected Service Water Battery cells and provides for battery restoration and maintenance, bas~d on the recommendation of IEEE Standard 450-1995. "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications,"

including the following:

1. Actions to restore battery cells with float voltage :5 2.13 V;

2. Actions to determine whether the float voltage of the remaining battery cells is >

2.13 V when the float voltage of a battery cell has been found to be.:5 2.13 V; i

3. Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates;

4. Limits on average electrolyte temperature, battery connection resistance, and battery terminal voltage; and

5. A requirement to obtain specific gravity readings of all cells at each discharge test, consistent with manufacturer recommendations.

OCONEE UNITS 1, 2, & 3 5.0-23 Amendment Nos. 386, 388, & 387

5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements Reporting Requirements 5.6 The following reports shall be submitted in accordance with 10 CFR 50.4.

5.6.1 Deleted 5.6.2 Annual Radiological Environmental Operating Report

N 0 T E ---------------------------------------------------

A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station.

The Annual Radiological Environmental Opdrating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year.

The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

OCONEE UNITS 1, 2, & 3 5.0-24 Amendment Nos. 386, 388, & 387

5.6 Reporting Requirements (continued) 5.6.3 Radioactive Effluent Release Report Reporting Requirements 5.6

.,-------------------------------NOTE.:--------------------------------------------------

A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

The Radioactive Effluent Release Report covering the operation of the unit in the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.

I The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR part 50, Appendix I,Section IV.B.1.

5.6.4 Deleted 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

Core operating limits shall be established, determined and issued in accordance with the following:

a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1.

Shutdown Margin limit for Specification 3.1.1;

2.

Moderator Temperature Coefficient limit for Specification 3.1.3;

3.

Physical Position, Sequence and Overlap limits for Specification 3.2J Rod Insertion Limits;

4.

AXIAL POWER IMBALANCE operating limits for Specification 3.2.2;

5.

QUADRANT POWER TILT (QPT) limits for Specification 3.2.3; OCONEE UNITS 1, 2, & 3 5.0-25 Amendment Nos. 386, 388, & 387

5.6 Reporting Requirements Reporting Requirements 5.6 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continded)

6.

Nuclear Overpower Flux/Flow/Imbalance and RCS Variable Low Pressure allowable value limits for Specification 3.3.1;

7.

RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits for Specification 3.4.1

8.

Core Flood Tanks Boron concentration limits for Specification 3.5.1;

9.

Borated Water Storage Tank Boron concentration limits for Specification 3.5.4;

10.

Spent Fuel Pobl Boron concentration limits for Specification 3.7.12;

11.

RCS and Transfer Canal boron concentration limits for Specification 3.9.1; and

12.

AXIAL POWER IMBALANCE protective limits and RCS Variable Low Pressure protective limits for Specification 2.1.1.

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

(1)

DPC-NE-1002-A, Reload Design Methodology II; (2)

NFS-1001-A, Reload Design Methodology; (3)

DPC-NE-2003-P-A, Oconee Nuclear Station Core Thermal Hydraulic Methodology Using VIPRE-01; (4)

DPC-NE-1004-A, Nuclear Design Methodology Using CASM0-3/SIMULATE-3P; (5)

DPC-NE-2008-P-A, Fuel Mechanical Reload Analysis Methodology Using TAC03 and GDTACO; (6)

BAW-10192-P-A, BWNT LOCA-BWNT Loss of Coolant Accident Evaluation Model for Once-Through Steam Generator Plants; OCONEE UNITS 1, 2, & 3 5.0-26 Amendment Nos.. 386, 388, & 387

5.6 Reporting Requirements Reporting Requirements 5.6 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

(7)

DPC-NE-3000-P-A, Thermal Hydraulic Transient Analysis Methodology; (8)

DPC-NE-2005-P-A, Thermal Hydraulic Statistical Core Design Methodology; (9)

DPC-NE-3005-P-A, UFSAR Chapter 15 Transient Analysis Methodology :

(10)

BAW-10227-P-A, Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel; I,

(11)

BAW-10164P-A, RELAP 5/MOD2-B&W-An Advanced Computer Program for Light Water Reactor LOCA and non-LOCA Transient Analysis; and (12)

DPC-NE-1006-P-A, Oconee Nuclear Design Methodology Using CASM0-4/SIMULATE-3 (Revision 0, May 2009).

The COLR will contain the complete identification for each of the Technical Specifications referenced topical reports used to prepare the COLR (i.e.,

report number, title, revision number, report date or NRC SER date, and any supplements).

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Eme__rgency Core Cooling System (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) and Main Feeder Bus Monitor Panel (MFPMP)

Report When a report is required by Condition B or G of LCO 3.3.8, "Post Accident Monitoring (PAM) Instrumentation" or Condition D of LCO 3.3.23, "Main Feeder Bus Monitor Panel," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring (PAM only),

the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

OCONEE UNITS 1, 2, & 3 5.0-27 Amendment Nos. 386, 388, & 387

5.6 Reporting Requirements 5.6.7 Tendon Surveillance Report Reporting Requirements 5.6 Any abnormal degradation of the containment structure detected during the tests required by the Pre-stressed Concrete Containment Tendon Surveillance Program shall be reported to the NRC within 30 days. The report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective action taken.

5.6.8 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5.5.10, St~am Generator (SG) Program. The report shall include:l

a.

The scope of inspections performed on each SG,

b.

Active degradation mechanisms found,

c.

Nondestructive examination techniques utilized for each degradation mechanism,

d.

Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e.

Number of tubes plugged during the inspection outage for each active degradation mechanism,

f.

Total number and percentage of tubes plugged to date,

g.

The results of condition monitoring, including the results of tube pulls and in-situ testing, and

h. The effective plugging percentage for all plugging in each SG.

OCONEE UNITS 1, 2, & 3 5.0-28 Amendment Nos. 386, 388, & 387

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 386 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-38 AMENDMENT NO. 388 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-47 AND AMENDMENT NO. 387 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-55 DUKE ENERGY CAROLINAS, LLC OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 DOCKET NOS. 50-269, 50-270, AND 50-287

1.0 INTRODUCTION

By appli<:;ation dated December 16, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML120030226), as supplemented by letters dated January 20, 2012 (ADAMS Accession No. ML12025A124), March 1, 2012 (ADAMS Accession No. ML12080A199), March 16, 2012 (ADAMS Accession No. ML12081A126), April18, 2012 (ADAMS Accession No. ML12110A175), July 11,2012 (ADAMS Accession No. ML12195A325),

July 20, 2012 (ADAMS Accession No. ML12207A109), August 31, 2012 (ADAMS Accession No. ML12249A400), November 2, 2012 (ADAMS Accession No. ML12312A031), April5, 2013 (ADAMS Accession No. ML13123A159), June 28, 2013 (ADAMS Accession No. ML13190A016),

August 7, 2013 (ADAMS Accession No. ML 1 ~228A268), December 18, 2013 (ADAMS Accession No. ML13358A042), February 14, 2014 (ADAMS Accession No. ML14055A068), April 3, 2014 (ADAMS Accession No. ML14099A264), April11, 2014 (ADAMS Accession No. ML14107A034),

and July 24, 2014 (ML14210A356), Duke Energy Carolinas, LLC (Duke Energy, the licensee),

submitted a license amendment request (LAR) for the Oconee Nuclear Station (ONS), Units 1, 2, and 3, to the U.S. Nuclear Regulatory Commission (NRC). The amendment would revise the ONS Technical Specifications (TSs) and the Updated Final Safety Analysis Report (UFSAR) to add the new Protected Service Water (PSW) System to the plant's licensing basis as an additional method of achieving and maintaining safe shutdown of the reactors in the event of a high-energy line break (HELB) or a fire in the shared Turbine Building.

Previously, by letter dated June 26, 2008 (ADAMS Accession No. ML081910559), the licensee requested an amendment to the operating licenses for ONS Units 1, 2, and 3. That LAR contained a complete revision of the licensing basis regarding the mitigation of HELB events

\\ outside of containment. That LAR was supplemented by letters dated December 22, 2008 (ADAMS Accession No. ML090020355) and June 29, 2009 (ADAMS Accession No. ML091870501 ). That LAR described the installation of a new system, the PSW system, intended, in part, to mitigate HELB events outside of containment. The PSW system provides a diverse means to achieve and maintain safe shutdown by providing secondary side decay heat removal (DHR), RCP seal cooling, RCS primary inventory control, and RCS boration for reactivity management following scenarios that disable the 4160 V essential electrical power distribution system. The PSW system is designed as a standby system for use under emergency conditions.

The PSW system provides added "defense-in-depth" protection by serving as a backup to existing safety systems and, as such, the system is not required to comply with single failure criteria. The PSW system is provided as an alternate means to achieve and maintain safe shutdown conditions for one, two, or three units following postulated scenarios that damage essential systems and components normally used for safe shutdown. The PSW system is not an Engineered Safety Feature Actuation System (ESFAS) and is not credited to mitigate design basis events as described in UFSAR Chapters 6 and 15. No credit is taken in the safety analyses for PSW system operation following design basis events.

The NRC staff has separated its review of the PSW system from the HELB licensing basis reconstitution LAR. By letter dated December 20, 2012 (ADAMS Accession No. ML12354A272),

the NRC staff informed Duke Energy of the suspension of its review of the balance of the HELB amendment request, until the NRC has issued a license amendment approving the final configuration of the PSW system.

By letter dated December 29, 2010 (ADAMS Accession No. ML103630612), the NRC staff issued license amendments for ONS Units 1, 2, and 3, approving the transition to a fire protection program in accordance with 10 CFR 50.48(c). Section 3.4.6 of that NRC safety evaluation (SE) credited the future installation of the PSW system for contributions to an overall reduction in the cumulative fire risk. The amendment revised ONS license condition 3.D for fire protection.

Transition License Condition 1 of the revised License Condition 3. D required completion of the items listed in Section 2:9, Table 2.9-1, of the NRC SE prior to January 1, 2013. Completion of the PSW system modifications is included in the requirements of that license condition and is a necessary step towards achieving full compliance with 10 CFR 50.48(c). The licensee did not meet the deadline for PSW system implementation specified in Transition License Condition 1 of the fire protection amendment; therefore, on July 1, 2013 (ADAMS Accession No. ML13224A919), the NRC staff issued a Confirmatory Order requiring Duke,Energy to complete PSW system implementation in accordance with the milestones and schedule specified in that order. Milestone's 1-3 of the Confirmatory Order required the licensee to provide the capability to supply offsite electrical power to the PSW building switchgear and from there to the Standby Shutdown Facility (SSF) switchgear; the capability to supply electrical power from each of the Keowee Hydro Units to the PSW building switchgear and from there to the SSF switchgear; and the capability to supply electrical power from the PSW building switchgear to simultaneously operate at least one high pressure injection pump per unit, and to operate the associated valves needed to align water flow to the reactor coolant pump seals and to inject water into the reactor coolant system. The licensee has notified the NRC of the completion of each of these 3 milestones, which were implemented under 10 CFR 50.59; consequently, the as-installed PSW electrical cable configurations are not within the scope of this license amendment.

This SE provides the technical bases for the staff's approval of the changes to the ONS licensing basis within the scope of these amendments. These amendments and the related SE do not approve nor endorse the as-installed PSW electrical system cable configurations. Duke Energy analyzed the PSW electrical system configurations and installed the PSW electrical cables and power supplies under the provisions of 10 CFR 50.59, and thus those parts of the system were not included in the scope of the staff's review for these amendments. The installed configurations of the PSW cabling and associated onsite power supply systems are the subject of a pending NRC inspection activity, as documented as an unresolved item in the NRC Component Design Basis Inspection Report, dated June 27, 2014, Section 1.2.b.v. (ADAMS Accession No. ML14178A535). In addition, Duke Energy submitted Licensee Event Report 269/2014-01, dated May 27, 2014 (ADAMS Accession No. ML14149A476), identifying an unanalyzed condition associated with the 13.8 kV emergency power cables located in an underground trench that also includes the PSW system 13.8 kV power cables.

The HELB licensing basis reconstitution LAR was supplemented by various letters subsequent to its acceptance for review. Information regarding the design of the PSW systerr) was also included in correspondence bet~een the NRC and the licensee regarding the licensing basis reconstitution for tornado protection and the transition to a risk-informed fire protection program. NRC review of the tornado protection licensing basis reconstitution LAR has also been deferred pending completion of the PSW amendment. The fire protection amendment was approved by the NRC staff as discussed above. Information provided by the licensee concerning the design of the PSW system has been updated and supplemented to reflect.information related to the HELB and tornado protection LARs and the fire protection amendment. These letters have been referenced where applicable.

The supplemental letters dated January 20, March 1, March 16, April 18, July 11, July 20, August 31, and November 2, 2012; April 5, June 28, August 7, and December 18, 2013; and February 14, April 3, April11, and July 24, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's proposed no significant hazards consideration determination as published in the Federal Register on JuJy 10, 2012 (77 FR 40652).

2.0 REGULATORY EVALUATION

The NRC staff considered a number of regulatory requirements and guidance documents in its review of the proposed License Condition, TSs and UFSAR changes for the PSW system for ONS Units 1, 2, and 3. The PSW is a new system providing significant improvement to the risk profile for the site. Additionally, the PSW system is not safety-related, and no credit is taken for its operation in any UFSAR Chapter 15 accident analyses. As such, the NRC staff considered certain requirements, standards, and guidance beyond the current licensing basis for ONS as part of its review criteria for this LAR. Those criteria used by the NRC staff in its review are discussed in this section.

2.1 General The principal design criteria for ONS were developed in consideration of the draft General Design Criteria (GDC) proposed by the Atomic Energy Commission (AEC) on July 11, 1967 (32 FR 10213; ADAMS Accession No. ML043310029). The ONS UFSAR, Section 3.1, contains a discussion of the ONS design criteria with regard to each of the draft GDC. The current GDC are generally comparable to the applicable AEC-proposed draft GDC, although the numbering is not consistent between the two.

Draft GDC 1, Quality Standards, requires that systems and components that are essential to the prevention of accidents be identified and then designed, fabricated, and erected to quality standards that reflect the importance of the safety function. As discussed in ONS UFSAR Section 3.1.1, the licensee has established a quality assurance program to satisfy the requirement of draft GDC 1 and the requirements of Appendix B to 10 CFR Part 50. Within this quality assurance program, essential systems and components are classified as QA-1 and are subject to the requirements of that classification.

Draft GDC 4, Sharing of Systems, states, "reactor facilities shall not share systems or components, unless it is shown safety is not impaired by the sharing." ONS UFSAR Section 3.1.4 addresses this criterion.

1 2i.2 lnservice Testing I

As required by 10 CFR 50.55a(f), inservice testing (1ST) of certain American Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with the requirements of ASME Operation and Maintenance Code for Nuclear Power Plants (OM Code) and applicable addenda, except where alternatives have been authorized or relief has been requested and granted by the Commission pursuant to Sections (a)(3)(i), (a)(3)(ii),

or (f)(6)(i) of 10 CFR 50.55a. 10 CFR 50.55a(f)(5)(ii) requires that, if a revised 1ST program for a facility conflicts with the TS for that facility, the licensee shall apply to the Commission for amendment of the TS to conform the TS to the revised program.

Subsection ISTB, "lnservice Testing of Pumps in Light-Water Reactor Power Plants," of the ASME OM Code establishes the requirements for 1ST of certain pumps used in nuclear power plants. Paragraph ISTB 1.1, "Applicability," states that pumps covered are those, provided with an emergency power source, that are required in shutting down a reactor to the safe shutdown condition, maintaining the safe shutdown condition, or mitigating the consequences of an accident. Subsection ISTC, "lnservice Testing of Valves in Light-Water Reactor Power Plants," of the ASME OM Code establishes the requirements for 1ST of certain valves and pressure relief devices used in nuclear power plants. Paragraph ISTC 1.1, "Applicability," states that valves covered are those that are required to perform a specificfunction in shutting down a reactor to the safe shutdown condition, maintaining the safe shutdown condition, or mitigating the consequences of an accident.

The licensee's 1ST program is based on the requirements of the 2004 Edition through 2006 Addenda of the ASME OM Code. The lice~see's 1ST program covers the fifth 1 0-year interval that started on July 1, 2012 and is currently scheduled to end on June 30, 2022. The licensee has indicated that the pumps and valves of the new PSW system will be included in the ONS 1ST program.

2.3 Review of PSW Safety Basis for High Energy Line Breaks The reactor safety performance requirements for the PSW system are established to ensure that the PSW provides a diverse method of protection following the postulated design basis HELB.

The NRC staff reviewed the licensee's analyses for the double-ended main steam line break (MSLB) and feedwater system pipe breaks in evaluating the PSW safety basis.

Double-Ended Main Steam Line Break The double-ended MSLB with loss of the essential power system is the limiting overcooling event which challenges the PSW system. The steam release from a rupture of a main steam pipe will result in an increase in steam flow, a reduction of coolant temperature and pressure, and an increase in core reactivity. The core reactivity increase may cause shutdown margin (SDM) to be lost. A return to power following a steam pipe rupture is a concern primarily b~cause of the high power peaking factors that would exist assuming the most reactive rod cluster control assembly to be stuck in its fully withdrawn position. The Reactor Protection System (RPS) and safety systems are actuated to mitigate the transient. The core is shut down by boric acid injection into the Reactor Coolant System (RCS) by the safety injection system. The rupture of a major steam line is the most-limiting cooldown transient. Decay heat would partly offset the cooldown, and reduce the post-trip return to power.

I The NRC staffs-review for the double-ended MSLB covered the following topics: (1) postulated initial core and reactor conditions, (2) methods of thermal and hydraulic analyses, (3) postulated sequence of events, (4) assumed responses of the reactor coolant and auxiliary systems, (5) functional and operational characteristics, (6) operator actions required to secure and maintain the reactor in a safe shutdown condition, (7) core power excursion, and (8) variables influencing neutronics.

Postulated accidents are equivalent to American Nuclear Society (ANS) Condition IV events, and anticipateq operational occurrences (AOOs) are equivalent to ANS Condition II and Ill events, as described in the NRC's Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (NUREG-0800), Section 15.0. Although this event is classified as a postulated accident as described in the SRP, it is analyzed to meet the acceptance criteria for an AOO, as indicated in Duke Energy's July 11, 2012, response to Request for Additional Information (RAI) 153 (ADAMS Accession No. ML12195A325). The acceptance criteria for AOOs are based on critical heat flux (CHF) not being exceeded, which the licensee demonstrates by analyzing the minimum departure from nuclear boiling ratio (DNBR) to determine that it remains above previously established limits. Specific review criteria considered by the staff are found in SRP Section 15.1.5.

The NRC's acceptance criteria for this portion of the review were based on the ONS current licensing basis and draft GDCs, as described in UFSAR Section 3.1 and discussed in Section 2.1 of this SE. Aspects of the following current GDCs were considered by the staff in its review of the PSW system; however, consideration of these GDCs does not revise the ONS licensing basis:

I (1) GDC-1 0, insofar as it requires the availability of instrumentation to monitor variables and systems over their anticipated ranges to assure adequate safety, and of appropriate controls to maintain these variables and systems within prescribed operating ranges; (2) GDC-17, insofar as it requires that an onsite and offsite electrical power system be provided to permit the functioning of the structures, systems, and components important to safety. The safety function for each system (assuming the other system is not functioning) shall be to provide sufficient capacity and capability to ensure that the acceptable fuel design limits and the design conditions of operational occurrence and that core cooling, containment integrity, and other vital functions are maintained in the event of an accident; (3) GDC-27 and GDC-28, insofar as they require that the reactor coolant system being designed with appropriate margin to ensure that acceptable fuel design limits are not exceeded and that the capability to cool the core is maintained; (4) GDC-31, insofar as it requires that the reactor coolant system be designed with sufficient margin to ensure that the boundary behaves in a non-brittle manner and that the probability of propagating fracture is minimized; (5) GDC-35, insofar as it requires the reactor coolant system and associated auxiliaries being designed to provide abundant emergency core cooling.

Feedwater System Pipe Breaks in the Turbine Building I

Depending upon the size and location of the break and the plant operating conditions at the time of the break, a feedwater pipe break could cause either an RCS cooldown (by releasing an excessive amount of energy through the break) or an RCS heatup (by reducing feedwater flow to the affected steam generator). Reactor protection and safety systems are available, and actuated to mitigate the event. The event for the purposes of establishing the reactor safety performance requirements of the PSW system is treated as an RCS heatup event.

The NRC staff's review for feedwater system pipe break events in the Turbine Building covered:

(1) postulated initial core and reactor conditions, (2) the methods of thermal and hydraulic analyses, (3) the sequence of events, (4) the predicted response of the reactor coolant and auxiliary systems, (5) the functional and operational characteristics of the reactor protection system, (6) operator actions, and (7) the results of the accident analyses.

The feedwater line break (FWLB) is classified as an ANS Condition IV event, or a postulated accident. The licensee analyzes the immediate effects of the event to confirm adherence to the more restrictive acceptance criteria of a Condition II event, which include the following:

(1) Pressure in the reactor coolant and main steam systems should be maintained below 110 perce.nt of the design values in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code.

(2) Fuel cladding integrity shall be maintained by ensuring that the minimum DNBR remains above the 95/95 DNBR limit.

(3) An AOO should not generate a postulated accident without other faults occurring independently or result in a consequential loss of function of the RCS or reactor containment barriers.

As a postulated accident, the FWLB analysis is also reviewed according to these criteria:

(1) Pressure in the RCS and main steam system should be maintained below acceptable design limits,.considering potential brittle as well as ductile failures.

(2) Fuel cladding integrity will be maintained if the minimum DNBR remains above the 95/95 DNBR limit. If the minimum DNBR does not meet these limits, then the fuel is assumed to have failed.

(3) The release of radioactive material shall not result in offsite doses in excess of the guidelines of 10 CFR Part 100.

(4) A postulated accident shall not, by itself, cause a consequential loss of required functions of systems needed to cope with the fault, including those of the RCS and the reactor containment system.

By adhering to these criteria, the licensee demonstrates that the reactor core and reactor coolant pressure boundary remain in an acceptable state following the postulated HELB until PSW flow can be established to the stealljl generators ( 15 minutes following the event), and High J Pressure Injection (HPI) flow to the RCP seals can be.established (20 minutes following the event).

The NRC staff's review was based on the guidance contained in SRP Section 15.2.8, acknowledging that the analyses support the reactor safety performance requirements for the PSW system, and that FWLBs are not in the Oconee licensing basis, as discussed in Chapter 15 of the ONS UFSAR.

2.4 Electrical Design The review criteria used by the NRC staff in its review of the PSW system electrical design included the following:

General Criteria for Nuclear Power Plant Construction Permits, Federal Register; July 11, 1967, Criterion 39-Emergency Power For Engineered Safety Features (Category A)- The plant auxiliary power distribution system shall be provided with an emergency power source under the control of the operator and shall be designed with sufficient redundancy and capacity to permit the functioning of the engineered safety features required to avoid undue risk to the health and safety of the public.

GDC 17, "Electric power systems," of Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50 requires, in part, that nuclear power plants have onsite and offsite electric power systems to permit the functioning of structures, systems, and components (SSCs) that are important to safety. The onsite system is required to have sufficient independence, redundancy, and testability to perform its safety function, assuming a single failure.

10 CFR 54.37(b) states, in part, that after the renewed license is issued, the FSAR update required by 10 CFR 50.71(e) must include any SSCs newly identified that would have been subject to an aging management review or evaluation of time-limited aging analyses in accordance with § 54.21. This FSAR update must describe how the effects of aging will be managed such that the intended function(s) in§ 54.4(b) will be effectively maintained during the period of extended operation.

RG 1.32, Revision 2, "Criteria for Safety-Related Electric Power Systems For Nuclear Power Plants" Section C. Regulatory Position b, states in part, "the battery charger capacity should be based on the largest combined demands of the various steady-state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the plant during which these demands occur." Regulatory Position c, states in part, that the battery service test should be performed in addition to the battery performance discharge tests.

RG 1.128, Revision 2, February 2007, "Installation Design and Installation of Vented Lead-Acid Storage Batteries for Nuclear Power Plants," which endorses (with certain clarifying regulatory positions described in Section C of this guide) IEEE Std 484-2002, "IEEE Recommended Practice for Installation Design and Installation of Vented Lead-Acid Batteries for Stationary Applications."

RG 1---:212, November 2012, "Sizing of large lead-acid storage batteries" This regulatory guide endorses (with certain clarifying regulatory positions described in Section C of this guide) IEEE Std. 485-1997, "IEEE Recommended Practice for Sizing of Lead-Acid Batteries for Stationary Applications."

I I

I RG 1.218, April2012, "Condition Monitoring Techniques for Electric Cables Used in Nuclear Power Plants," Section C, Regulatory Position 2, states in part that the NRC staff considers the use of appropriately selected combinations of typical cable condition-monitoring techniques, such as those discussed in Section B of the RG, within the framework of a comprehensive cable condition-monitoring program to be an acceptable method for satisfying the Commission's regulations to assess the continuity of the systems and the conditions of their components. The condition-monitoring techniques selected should be based on plant-specific design, installation, and operating conditions and operating experience related to the cables used in nuclear plants.

2.5 Mechanical Design The review criteria considered by the NRC staff in its review of the structural integrity of the mechanical portion (piping and heating, ventilation, and air conditioning (HVAC)) of the PSW system are based on 10 CFR 50.55a and the Design Criteria specified in the ONS UFSAR, since ONS was not licensed to the 10 CFR 50, Appendix A, General Design Criteria. The ONS Design Criteria are described and compared to the corresponding 10 CFR 50, Appendix A, GDC below.

Specific review guidance also used during this evaluation is contained in NUREG-0800, SRP, Sections 3.9.1, 3.9.2, 3.9.3, and 5.2.1.1, Regulatory Guides (RGs) 1.29, 1.52, and 1.140, Revision 2, American Institute of Steel Construction (AISC) Manual, 61h Edition, 1963, and the ASME AG-1.

10 CFR 50.55a requires that safety-related pressure-retaining components of fluid systems meet applicable ASME code requirements.

GDC 1 requires,.in part, that SSCs important to safety be designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety functions to be performed.

ONS Criterion 1 requires that those systems and components of reactor facilities which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be identified and then designed, fabricated, and erected to quality standards that reflect the importance of the safety function to be performed. Where generally recognized codes or standards on design, materials, fabrication, and inspection are used, they shall be identified. Where adherence to such codes or standards does not suffice to

( assure a quality product in keeping with the safety function, they shall be supplemented or modified as necessary.

GDC 2 requires, in part, that SSCs important to safety be designed to withstand the effects of earthquakes combined with the effects of normal or accident conditions.

ONS Criterion 2 requires that those systems and components of reactor facilities which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be designed, fabricated and erected to performance standards that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects.

GDC 4 require:s, in part, that SSCs important to safety b~ designed to qccommodate the effects of and be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents.

ONS Criterion 40 requires that protection for engineered safety features shall be provided against dynamic effects and missiles that might result from plant equipment failures.

2.6 Structural Design The review criteria considered by the NRC staff in its review of the structural integrity of the PSW Building and the associated underground electrical duct banks are based on information contained in the UFSAR, including the ONS design criteria.

Section 3.2.1.1.1 of the ONS UFSAR describes Class 1 structures as those SSCs which prevent uncontrolled release of radioactivity and are designed to withstand all loadings without loss of function.

UFSAR Section 3.1.2, "Criterion 2-Performance Standards," of the ONS UFSAR states the following:

Those systems and components of reactor facilities which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be designed, fabricated and erected to performance standards that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects.

The ONS UFSAR, Section 3.3.2 indicates that Revision 1 of Regulatory Guide (RG) 1.76, "Design-Basis Tornado and Tornado Missiles for Nuclear Power Plants," was incorporated into the plant's licensing basis and the new systems (and their associated components and/or structures) that are required to resist tornado loading will conform to the tornado wind, differential pressure, and missile criteria specified in RG 1. 76, Revision 1.

Other regulatory guidance documents that were considered in the review of the structural design of the PSW Building and the associated underground electrical duct banks are as follows:

NUREG-0800, Standard "Review Plan (SRP), Section 3.8.4 "Other Seismic Category 1 Structures,'" provides acceptance criteria and cites appropriate regulatory guidance for the design and construction of safety-related structures to withstand the effects of natural

phenomena without loss of capability to perform their intended safety function.

SRP 3.8.5, "Foundations," provides acceptance criteria for the design of foundation for safety-related structures, including factors of safety against overturning, sliding, and flotation to ensure adequate safety margins.

American Concrete Institute (ACI) 349 "Code Requirements for Nuclear Safety Related Concrete Structures," and American National Standardllnstitute (ANSI)/American Institute of Steel Construction (AISC) N690 "Specification for the Design, Fabrication and Erection of St~el Safety Related Structures for Nuclear Facilities," co~tain basic requirements and load combinations for the design of safety related concrete and steel structures, respectively.

RG 1.142 "Safety-Related Concrete Structures for Nuclear Power Plants," Revision 2 endorses ACI 349-97 with exceptions.

RG 1.199 "Anchoring Components and Structural Supports in Concrete," dated November 2003, provides guidance to the licensees on methods acceptable to the NRC staff for complying with the NRC's regulations in the design, evaluation, and quality assurance of anchors used for component and structural supports. RG 1.199 endorses Appendix B to ACI 349-01, with exceptions.

SRP Section 3.7.1, "Seismic Design Parameters," and Section 3.7.~, "Seismic System Analysis," provide guidance relative to calculating seismic responses of safety-related structures, structural modeling, and dynamic analysis method (response spectrum analysis or

.time history analysis) that are acceptable to the NRC staff.

RG 1.122 "Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components," Revision 1 (February 1978) describes methods generally acceptable to the NRC staff for developing the horizontal and the vertical in-structure response spectra (floor response spectra) from the time history motions resulting from the dynamic analysis.

RG 1.92 "Combining Modal Responses and Spatial Components in Seismic Response Analysis," Revision 2 (July 2006) describes methods acceptable to the NRC staff (1) for combining the responses due to three components of earthquake motion, for both the response spectrum method and the time history method; and (2) for combining of modal responses, including consideration of closely-spaced modes and high-frequency modes, when the response spectrum method of analysis is used to determine the dynamic response of damped linear systems.

2.7 Human Factors The regulatory requirements and guidance which the NRC staff considered in assessing the human factors aspects of the proposed PSW amendment are as follows:

10 CFR 50.120, "Training and qualification of nuclear power plant personnel."

NUREG-0800, SRP, Chapter 18, "Human Factors Engineering," provides review guidance regarding the requirements for an effective hum*an factors engineering design NUREG-0711, "Human Factors Engineering Program Review Model," Revision 3, provides acceptance criteria for human factors engineering program safety evaluations.

2.8 Technical Specifications The PSW system is provided as an alternate means to achieve and maintain safe shutdown conditions for one, two, or three units following postulated scenarios that damage essential systems and components normally used for safe shutdown. Based on its contribution to the reduction of overall plant risk, the PSW system satisfi~s Criterion 4 of 10 CFR 50.36 (c)(2)(ii) and is therefore included in the TSs. As part of its amendment request, the licensee proposed the addition ofTS 3.7.10, "Protected Service Water (PSW) System," TS 3.7.10a, "Protected Service Water (PSW) Battery Cell Parameters," and TS 5.5.22, "Protected Service Water System Battery Monitoring and Maintenance Program."

Section 182a of the Atomic Energy Act (the "Act") requires applicants for nuclear power plant operating licenses to include TS as part of the application. The TS ensure the operational capability of SSCs that are required to protect the health and safety of the public. The Commission's regulatory requirements related to the content of the TS are contained in 10 CFR Section 50.36. That regulation requires that the TS include items in the following specific categories: (1) safety limits, limiting safety systems settings, and limiting control settings (50.36(c)(1 )); (2) Limiting Conditions for Operation (50.36(c)(2)); (3) Surveillance Requirements (50.36(c)(3)); (4) design features (50.36(c)(4)); and (5) administrative controls (50.36(c)(5)).

10 CFR 50.36(c)(2) states that LCOs are the lowest functional capability or performance level of equipment required for safe operation of the facility and when LCOs are not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TS until the LCO can be met.

10 CFR 50.36(c)(2)(ii) states:

A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

(A) Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

(B) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(C) Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

\\

(D) Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

10 CFR 50.36(c)(3) states that SRs are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

In general, there are two classes of changes toTS: (1) Changes needed to reflect modifications to the design basis (TS are derived from the design basis), and (2) voluntary changes to take advantage of the evolution in policy and guidance as to the required content and preferred format of TS over time. This amendment addresses the first class of changes.

The NRC's guidance for the format and content of licensee TSs can be found in NUREG-1430, "Standard Technical Specifications Babcock land Wilcox Plants."

3.0 TECHNICAL EVALUATION

The NRC staff's technical evaluation of the PSW license amendment request is documented in the following 8 sections of this SE: Plant Systems, lnservice Testing, Reactor Safety, Electrical, Mechanical/Structural, Seismic, Human Factors, and Technical Specifications.

3.1 Plant Systems Review 3.1.1 Protected Service Water System Description The licensee proposed the installation of a new PSW system. The PSW system provides an alternate means to achieve and maintain safe shutdown conditions for one, two, or all three units at ONS. The PSW system is intended to mitigate postulated events resulting in damage to equipment in the common Turbine Building and the loss of existing essential systems and components credited to achieve and maintain safe' shutdown conditions.

By letter dated December 18, 2013, (ADAMS Accession No. ML13358A042), as part of the amendment request, the licensee submitted proposed ONS UFSAR Section 9.7, a new UFSAR section that describes the PSW system design and functions. This submittal replaced previously submitted UFSAR Section 9.7 descriptions.

The electrical portion of the PSW system provides power to PSW components and backup power to components of other systems that are needed to achieve and maintain safe shutdown conditions. In addition, the PSW system 'is capable of providing backup electrical power to the SSF for defense-in-depth. Major components of the PSW electrical system are located in the PSW Building, a reinforced concrete structure. Power for the PSW system is supplied by a connection from the Central Tie Switchyard (a non-QA 1 source), or by an underground connection to the Keowee Hydroelectric Units (KHUs, a QA 1 source).

The mechanical portion of the PSW system consists of one high-head pump and one booster pump in combination, which take suction from the Unit 2 Condenser Circulating Water (CCW) embedded piping and supply raw Lake KeoiNee water to the secondary side of the steam generators (SGs) of all3 units. The PSW pumps are capable of supplying water to the SGs of all units via the Emergency Feedwater (EFW) headers. The lake water is vaporized in the SGs and then released to the atmosphere to remove decay heat from the RCS.

The PSW system provides backup power to a HPI pump and associated HPI valves in each unit.

The HPI pump supplies borated water from the Borated Water Storage Tank (BWST) to the RCS.

This accomplishes the design bases of reactivity control, RCP seal cooling, and RCS water level recovery. The PSW system provides backup power to the RCS and reactor vessel head high point vent valves for reactivity control and RCS inventory control. Backup power can also be provided to select groups of pressurizer heaters for RCS pressure control if the RCS water level is sufficient for pressurizer heater operation.

As described in UFSAR Section 9.7.2, the design criteria for the PSW system are as follows:

Major PSW components ~re Duke Energy Quality Assurance Condition 1 (QA-1 ).

Components that receive backup power from PSW or systems that connect to PSW retain their existing seismic and quality classifications.

Maintain a minimum water level above the reactor core and maintain Reactor Coolant Pump Seal cooling. In addition, maintains RCS subcooling for fire scenarios.

Provide steam generator secondary side cooling water from Lake Keowee to promote natural circulation core cooling.

Transfer decay heat from the RCS by steaming the SGs to atmosphere.

Maintain Keff < 0.99 after all normal sources of RCS makeup have become unavailable, by providing makeup via the HPI system which supplies makeup of a sufficient boron concentration from the BWSTs.

'I Control of PSW primary and booster pumps, motor operated valves and solenoid valves, required to bring the system into service are controlled from the Main Control Rooms (MCRs ).

3.1.2 Evaluation of Proposed Changes 3.1.2.1 Single Failure Criterion The LAR states that the PSW system is a defense-in-depth backup to existing safety-related systems and components. As such, it was not designed to withstand a single active failure. The NRC staff agrees that the application of the. single failure criterion is not required for defense-in-depth systems. However, the PSW system may be relied upon to achieve and maintain safe shutdown conditions following the loss of essential systems and components in the Turbine Building. In such an event, although the single failure criterion does not apply, the availability of the PSW system together with the SSF provides enhanced capability to ensure safe shutdown.

Currently the SSF, described in ONS UFSAR Section 9.6, provides one train of standby shutdown equipment for use if essential systems and components located in the Turbine Building are unavailable. The PSW system is designed to provide an additional train of standby shutdown equipment independent from the SSF. Together, the PSW system and SSF provide diversity of safe shutdown equipment independent of the essential systems and components located in the Turbine Building. These two systems are independent with the exception of a common water source for secondary side decay heat removal.

The licensee's strategy for treating the PSW system and SSF as two diverse systems for the application of single active failures was documented in a letter to the NRC dated November 30, 2006 (ADAMS Accession No. ML070290328). A postulated single active failure in the PSW system or SSF, with the exception of failures in the common water source, does not prevent the other system from achieving and maintaining safe shutdown conditions. During a meeting with the licensee on March 5, 2007, (meeting summary dated March 28, 2007, ADAMS Accession No. ML070750126) the NRC staff provided the results of a pre-application review of the ONS strategies for tornado and HELB mitigation (ADAMS Accession No. ML070530070). The NRC staff stated that the combined use of the PSW system and SSF was a viable strategy for event mitigation and identified additional issues that required clarification in the forthcoming application.

These issues concerned the tornado and HELB mitigation strategies and licensing bases reconstitution, which are the subjects of separate NRC reviews that have been suspended pending the completion of this review, and therefore, they are not addressed in this SE.

The NRC staff agrees that the PSW system, as a defense-in-depth addition to ONS, is not required to withstand a single active failure. For those scenarios where the PSW system is relied upon to achieve and maintain safe shutdown conditions in the reactors, the SSF is considered to be an additional train of safe shutdown equipment.

3.1.2.2 Quality Assurance Standards 1

As described in ONS UFSAR Section 3.1.1.1, the Quality Assurance (QA)-1 program satisfies the requirement of draft GDC 1 and 10 CFR 50 Appendix B. The QA-1 program is applied to "essential systems and components" as identified in the original licensing basis. Other systems and components, including portions of the SSF, have been included in the QA-1 program subsequent to the original licensing as a result of commitments made by the licensee.

The licensee has stated in the LAR that major components of the PSW system will be included in the QA-1 program. This establishes quality assurance standards that are comparable to those for ONS essential systems and components. While the PSW system is not classified as an essential system, it may be relied upon to achieve and maintain safe shutdown conditions following a postulated loss df essential systems and components located in the Turbine Building. The NRC staff finds that the inclusion of the PSW system in the QA-1 program establishes a level of quality assurance commensurate with the system's safety significance. Existing systems and components that interface with the PSW system retain their current quality and seismic pedigree.

3.1.2.3 RCP Seal Cooling, RCS Makeup, and Reactivity Control Following the postulated loss of essential systems and components in the Turbine Building, the PSW system provides backup electrical power to one train of the HPI system per unit.

Additionally, the PSW system provides backup power to the RCS high point vent valves, reactor vessel head high point vent valves, and select groups of pressurizer heaters. Propose~ ONS UFSAR Section 9.7 describes how these components will be used to provide RCP seal cooling, RCS inventory control, reactor core reactivity control, and RCS letdown.

The HPI system is a portion of the broader Emergency Core Cooling System (ECCS) which is designed to cool the reactor core and provide shutdown capability following design basis accidents. During normal operation, the HPI system provides makeup to the RCS and seal injection water for the reactor coolant pumps. One HPI pump is capable of providing the required flow for RCP seal cooling, as stated in ONS UFSAR Section 9.3.2.2.2. In the emergency mode of operation, described in ONS UFSAR Section 6.3.2.2.1, the HPI pumps inject borated water from the BWST for RCS inventory makeup, RCP seal cooling, and reactivity control.

During PSW system operation, one HPI train per unit is aligned to the BWST and operated in the emergency mode.. In this mode of operation, the HPI pump injects borated water into the RCS and RCP seals. Th'e flow to the RCP seals is adequate to provide cooling. jln response to RAI 58 dated December 7, 1 2010, (ADAMS Accession No. ML103470388) the licensee stated that the transfer of these components to PSW electrical backup power can be accomplished from the control room within 20 minutes. The restoration of RCP seal cooling within this time frame is adequate to prevent degradation of the seals. One train of the HPI system is sufficient to meet the cooling requirement for the RCP seals and backup power from the PSW can be aligned within an acceptable time frame. Therefore, the NRC staff finds that the operation of the HPI pump for RCP seal cooling is acceptable.

Operation of the HPI pump requires a flow of cooling water to the HPI pump motor bearings. The HPI pump motor bearings are normally cooled by water from the Low Pressure Service Water (LPSW) system, which is assumed unavailable following the loss of essential systems and components in the Turbine Building. The PSW system is designed to provide 10 gpm of cooling water to the HPI pump motor bearings via a connection to the PSW header. In response to RAI 2-47 dated June 24, 2010, (ADAMS Accession No. ML101830011) the licensee clarified that the minimum acceptable flow rate for cooling water flow to the HPI pump motor bearings is 1.5 gpm/HPI pump. The design flow rate of 10 gpm/unit ensures the HPI pump motor bearings will be provided with adequate cooling water flow to support operation of the HPI pump. The NRC staff finds that the PS..W system design is adequate to support operation of one HPI train per unit.

3.1.2.4 Secondary Side Decay Heat Removal Following the postulated loss of essential systems and components in the Turbine Building, the PSW system is designed to perform the secondary side decay heat removal function. The PSW system utilizes one high-head pump and one booster pump to supply water from Lake Keowee to the secondary side of the SGs. The high-head pump and the booster pump are located in the Auxiliary Building and replace the existing Auxiliary Station Service Water (ASW) pump. Power for the PSW system pumps and the associated valves is supplied from the PSW electrical system.

The PSW high-head pump discharge is routed to the A and B SGs of Units 1, 2, and 3, via a connection to the EFW headers.

As stated in the system description, the PSW system must be capable of~ providing sufficient feedwater flow to the SGs to remove decay heat from the RCS. The NRC staff reviewed the proposed ONS UFSAR Section 9. 7 and the PSW system flow diagram provided by letter dated December 18, 2013 (ADAMS Accession No. ML13358A042). The PSW high-head pump is designed to supply lake water to the SGs at a rate of 375 gpm per unit at 1082 psig with the SGs at the lowest safety relief valve setpoint. During operation of the PSW high-head pump, check valves in the EFW header prevent backflow towards the inactive EFW pumps. The PSW system is provided with flow control valves that are operated manually from the control room. Sufficient instrumentation is provided in the control room with power from the PSW electrical system to monitor performance of the PSW system and adjust feedwater flow rates to each SG. Flow to a depressurized SG can be terminated from the control room to prevent overcooling the RCS.

Therefore, the NRC staff finds that the PSW system has been designed with the capability to provide sufficient feedwater flow to all units, and that the system has the capability to manually control feedwater flow as needed.

The water source for PSW must be able to support the system's secondary side decay heat removal function during system operation. The PSW system draws water from the Unit 2 CCW embedded. piping with makeup provided by a portable submersible. pump powered from the PSW or SSF sys~ems. The CCW system is designed to take advantage cpf a siphon effect to draw water from the lake, as described in ONS UFSAR Section 9.2.2.2.1. Siphon flow into the CCW system maintains the water inventory in the CCW embedded piping following a loss of power to the CCW pumps. If CCW forced flow and siphon flow are both lost, operator actions must be taken to put the portable submersible pump into service. In response to RAI 103 dated December 16, 2011 (ADAMS Accession No. ML12003A070), the licensee provided additional information regarding the timing of this action. In the worst-case scenario, the minimum time available to put the portable pump into service was 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months 20 minutes, following the loss of CCW forced flow and siphon flow. The licensee stated that a minimum time to deploy the portable submersible pump was previously determined in support of the SSF, and that the minimum time for that scenario bounded the worst-case scenario for the PSW system.

In order to support operation of the PSW system, the portable submersible pump must be capable of providing makeup flow rates that exceed the PSW system suction flow rate. The response to RAI 103 confirmed that the portable submersible pump is capable of providing makeup at a flow rate sufficient to support extended operation of the PSW system. In addition, the licensee stated in the response that an identical portable pump is available on site as a backup to the portable submersible pump stored in the SSF. While the PSW system is not designed to withstand a single active failure,Jthe availability of a replacement portable pump provides additional assurance that the suction source will be available. The NRC staff finds that the CCW system together with the portable submersible pump is capable of supplying adequate volume of lake water to support the PSW system's secondary side decay heat removal function.

Lake Keowee, via the CCW system intake canal and piping, serves the ultimate heat sink for ONS.

As described in ONS UFSAR Section 9.2.2.1, the CCW intake canal includes a submerged weir to create an emergency pond of cooling water should Lake Keowee be lost due to a dam failure.

In the event of a loss of access to cooling water from Lake Keowee with subsequent loss of the submerged weir, the volume of water that remains trapped in the CCW intake and discharge piping of the three ONS units is relied upon as the alternative heat sink. The PSW system replaces the ASW system for the purpose of delivering water from the alternative heat sink to the secondary side of the steam generators in this scenario. The PSW system is put into service after the EFW water source\\has been expended. Cross-connects between the Unit 2 CCW and the Units 1 and 3 CCW embedded piping are opened, which enables the PSW system to draw water from the combined volume of the three units' CCW piping by siphon effect. The PSW system pumps water from Unit 2 CCW system piping to the secondary side of the steam generators in the three ONS units. The PSW system has been designed to perform the secondary side decay heat removal function for a period of at least 30 days without exceeding the system temperature limit while maintaining pump net positive suction head requirements.

The current licensing basis for the ultimate heat sink following a loss of Lake Keowee and subsequent loss of the submerged weir is described in ONS UFSAR Section 3.2.2. The proposed revision to ONS UFSAR Section *3.2.2 establishes a minimum of 30 days cooling water supply with the PSW system in use. The PSW system can draw on the volume of water retained in the CCW piping of all three ONS units, which is sufficient to provide secondary side decay heat removal for at least 30 days with additional margin. Operation of the PSW system does not increase the temperature of the water in the CCW volume beyond the design limits for the PSW system components. While the licensee has not committed to RG 1.27, Ultimate Heat Sink for Nuclear Power Plants, the NRC staff took this guidance into consideration. A period of 30 days provides sufficient time to evaluate the situation and take corrective actions and conforms to the su:ggested ultimate heat sink capability expressed in Regulatory Position 1 of RG 1.27. Therefore, the NRC staff finds that the ultimate heat sink will provide sufficient cooling water for at least 30 days, with the PSW system in use, and that the proposed change to ONS UFSAR Section 3.2.2 is acceptable.

3.1..2.5 Replacement of the Station Auxiliary Service Water System The PSW pump will be installed in place of the Station ASW pump in the Auxiliary Building. The Station ASW discharge piping will be completely replaced by higher schedule piping to meet PSW pump discharge pressure design requirements. The NRC staff reviewed the ASW system to ensure that its replacement would not result in a decrease in safety.

As described in ONS UFSAR Section 9.2.3, the ASW system was designed to provide secondary side decay heat removal following a loss of the main feedwater system, EFW system, and decay heat removal system. The ASW pump supplies Lake Keowee water from the Unit 2 CCW

. embedded piping to the secondary side of the SGs. The ASW pump can provide adequate flow to*

remove decay heat in all three units. However, a manual action must be taken to depressurize the SGs before the ASW can be put into service. Power for the ASW pump is supplied from the standby bus number 1.

The PSW system also supplies Lake Keowee water from the Unit 2 CCW embedded piping to the secondary side of the SGs for decay heat removal. It is sized to provide adequate flow for secondary side decay heat removal to all three units concurrently at a pressure of 1082 psig. The higher discharge pressure of the PSW system means the SGs do not need to be depressurized prior to placing the PSW pump in service. Also, the PSW system is provided with an independent power source that is available in postulated scenarios where the standby bus is not available.

The PSW system design exceeds the capabilities of the ASW system. Additionally, the PSW system provides a more reliable means of performing the secondary side decay heat removal function. Therefore, the NRC staff finds the replacement of the ASW system with the PSW system to be acceptable.

3.1.2.6 Alternate Auxiliary Building and Reactor Building Cooling The PSW system and components must be capable of performing in the environment to which they are expected to be exposed. In response to RAI 75, dated December 16, 2011 (ADAMS Accession No. ML12003A070), the licensee stated that the specifications for the PSW system were based on the existing temperature profiles following a design basis accident as documented in the Environmental Qualification Criteria Manual. Calculations of the additional heat load contributed by PSW equipment were incomplete at that time, and the licensee stated that additional modifications would be made to provide adequate heat removal capability to maintain the Auxiliary Building temperature within the equipment specifications. Subsequently, the licensee determined that additional modifications would be required.

In response to RAI 170 dated June 28, 2013 (ADAMS Accession No. ML13190A016), the licensee provided a description of the strategy to restore cooling for the Auxiliary Building and Reactor Building and maintain temperatures within PSW equipment specifications. Cooling for the Auxiliary Building relies on the LPSW system and the Auxiliary Building Chilled Water (CW) system, which rely on equipment located in the Turbine Building. In order to restore cooling, a selection of existing Auxiliary Building exhaust fa~s and air handling units will be provided with backup electrical power from the PSW electrical system and a backup source of chilled water from a new system designated the Alternate Chilled Water (AWC) system. Reactor Building cooling is restored using a diesel driven pump to supply copling water to one Reactor Building Cooling Unit (RBCU) and supplying backup electrical power from the PSW electrical system.

The NRC staff requested additional information regarding how the AWC system design accomplished its intended function. In responseto RAI188 dated December 18, 2013, (ADAMS Accession No. ML13358A042) the licensee provided additional details of the AWC system design.

The AWC is a closed loop chilled water system that provides cooling to the air handling units in

. the Auxiliary Building via two headers. One AWC system header is used to supply air handling units normally cooled by LPSW and the other header supplies those cooled by the CW system.

The AWC system is not designed to withstand a single failure, but includes two full capacity chillers.

The licensee performed calculations to determine the temperature response of the auxiliary and reactor buildings during extended operation of the PSW system with the AWC system in service.

PSW system equipment ratings were then verified against the predicted temperatures to ensure that all equipment remained capable of operation. Similar calculations were performed for the reactor building in order to evaluate the design of the reactor building alternate cooling. The NRC staff reviewed the seeping analysis calculations used to establish the AWC system and reactor building alternate cooling design.

The licensee developed a Generation of Thermal-Hydraulic Information for Containments (GOTHIC) computer model to analyze the temperature response in the Auxiliary Building during the extended operation of the PSW system. The licensee used the GOTHIC version 8.0 computer code for the AB heat up analysis. GOTHIC 8.0 is the latest version and a state-of-the-art general purpose thermal-hydraulics computer code maintained by Numerical Applications, Inc. (NAI), for the Electric Power Research Institute (EPRI) for performing containment analyses. GOTHIC is qualified under the NAI Quality Assurance (QA) program which conforms to the requirements of 10 CFR Part 50 Appendix B, with error reporting in accordance with 1p CFR Part 21. GOTHIC is widely used by the nuclear industry and applications of this code have been previously approved by the NRC staff on a case-by-case basis.

The NRC staff evaluated the major assumptions used by the licensee in the GOTHIC model and found that they were appropriately conservative. Random consistency checks were performed between the reference material identified by the licensee and the GOTHIC calculation. These checks did not identify any inconsistencies between the inputs identified in the reference material and the GOTHIC calculation. Additionally, the NRC staff performed an execution of the GOTHIC calculation over a reduced transient time. The results of the NRC staff's execution of the GOTHIC calculation were consistent with the results provided by the licensee. The NRC staff evaluation determined that the GOTHIC analysis was acceptable and was performed in accordance with 10 CFR 50 Appendix B requirements. Therefore, the NRC staff finds that the use of this methodology acceptable.

The PSW system is designed to mitigate events resulting in damage to essential systems and components located in the Turbine Building normally used for safe shutdown. Therefore, the AWC system must not rely on componen~s located in the Turbine Building if it is to support the I

operation of the PSW system during a credited event. In response to RAI 189 dated December 18, 2013, (ADAMS Accession No. ML13358A042) the licensee stated that the "AWC piping, power and control cables, and equipment will be located outside of the Turbine Building envelope."

As a r~sult, the AWC system is not impacted by events initiating in the Turbine Building and can be assumed available to support the operation of the PSW system.

The AWC system and alternate reactor building cooling are capable of providing sufficient cooling to maintain the building temperatures within equipment specifications during PSW system operation. Operator actions to align these systems for operation will be captured in procedures and the timeframe to perform these actions has been determined. Equipment used for the AWC system and alternate reactor building cooling is located outside of the Turbine Building, such that it is available following events that result in damage to essential systems and components located in the Turbine Building. Therefore, the NRC staff finds the proposed addition of the AWC system and alternate reactor building cooling acceptable to support the extended operation of the PSW system.

3.2 lnservice Testing (1ST) of the PSW System 3.2.1

Background

The NRC staff reviewed the design and operation of the mechanical components of the PSW system, as described in the LAR, its supplements, and the proposed new UFSAR section. In letters dated February 16, 2012 (ADAMS Accession No. ML12047A177) and June 11, 2012 (ADAMS Accession No. ML12144A444 ), the staff issued RAis to the licensee regarding in service testing (1ST) of pumps and valves in the PSW system. In response to the NRC staff's RAis, the licensee provided additional information in letters dated March 16, 2012 (ADAMS Accession No. ML12081A126), and August 31, 2012 (ADAMS Accession No. ML12249A400).

3.2.2 Licensee's Proposed TS Changes and FSAR Updates The new PSW system is provided as an alternate means to achieve and maintain a stable RCS pressure and temperature for the Oconee Nuclear Station, Units 1, 2, and 3, following HELB or fire events. The system consists of one primary pump, one booster pump and one portable pump, and a number of valves that are required for PSW system operation. The safety function provided by the PSW system is to supply cooling water for decay heat removal at full system pressure to all six steam generators in the three units (two steam generators per unit) following certain postulated accident events. A secondary safety function of the PSW is, in conjunction with the High Pressure Injection System, to provide borated water to the RCS pump seals and to provide primary RCS makeup. For extended operation, the PSW portable pump is designed to provide a supply of backup water to the PSW system in the event of loss of CCW and subsequent loss of CCW siphon flow. As such, the PSW portable pump would serve a backup function to provide sufficient cooling water from Lake Keowee to allow the PSW system to maintain the reactor in safe shutdown condition for an extended period following those postulated accidents for which the normal safety systems might be unavailable.

The LAR included a newTS Section 3.7.10, "Protected Service Water System," and a new UFSAR Section 9.7 for the PSW system. Specifically, SR 3.7.10.3 is added to require the licensee to verify that the developed head of PSW primary and booster pumps is greater than or equal to the required developed ~ead at the flow test point at a frequency in accordance! with the lnservice Testing Program.

The design bases and safety function of the PSW system and associated components including pumps and valves are described in the proposed UFSAR Section 9.7. In the new PSW system, pumps, solenoid-operated valves, and motor-operated valves (MOVs) are used to move water and isolate the flow in the system. Check valves and manually-operated valves are used to prevent back-flow, accommodate testing, or are used for system isolation during maintenance.

All pumps and valves in the system are ASME Section Ill, Class 3.

3.2.3 Staff Evaluation of 1ST for the PSW System The NRC staff reviewed the licensee's proposed LAR as it relates to the 1ST of certain pumps and valves associated with the new PSW system, and determined that additional information was required in order to complete the review. Specifically, the licensee was requested to provide UFSAR updates and revisions related to PSW pumps and valves, and to provide proposed changes to the licensee's 1ST program, including the bases for excluding any PSW pumps and valves from the 1ST program, and modifications to the licensee's MOV program. The licensee was also asked to identify any requests to the NRC for alternatives or relief from the ASME OM Code requirements for the 1ST of PSW pumps and valves.

The licensee provided a new UFSAR section and updated pages addressing the PSW system modification. The NRC staff reviewed the new UFSAR Section 9.7, "Protected Service Water System," and updated pages, and found that the functions of the PSW pumps and valves were adequately described and that the UFSAR updates were acceptable.

The PSW system configuration includes a minimum flow line to facilitate inservice testing of the PSW pumps. In order to facilitate testing of the system flowpath to the SGs, the PSW pump can be aligned to an alternate source of condensate water. The NRC staff finds that the PSW system design includes adequate provisions for inservice testing of the pumps.

The licensee also provided a table depicting the 1ST-related components (i.e., pumps and valves),

in the PSW system; and stated that these components will be monitored within the Duke Energy 1ST Program. The licensee noted that all PSW vent, drain, and instrument isolation valves are not listed and are not in the licensee's 1ST program. Paragraph ISTC 1.2, "Exclusions" of the ASME OM Code excludes valves used only for operating convenience such as vent, drain, instrument, and test valves. Therefore, excluding vent, drain, and instrument isolation valves from the 1ST Program is consistent with paragraph ISTC 1.2 of the ASME OM Code, and is acceptable.

The PSW primary and booster pumps, but not the portable pump, are included in the same 1ST table noted above. The licensee noted in responses to questions from the NRC that a portable pump powered by either the PSW system or the SSF would be utilized to pump water from the intake canal into the Unit 2 CCW inlet piping to provide long-term inventory control in the CCW piping to support extended operation of the PSW pumps. The licensee also noted that the PSW portable pump will not be powered from an emergency power source. Paragraph ISTB 1.1 of the ASME OM Code excludes pumps that are not provided with an emergency power source. On the basis that the PSW portable pump is not required to be powered from an emergency power source, the NRC staff determined that it is acceptable to exclude this pump from the 1ST program.

I Additionally, the licensey noted that the active MOVs used for PSW operation ""!ill be included in the ONS Generic Letter (GL) 89-10 MOV Program, which encompasses GL 89-10 "Safety-Related MOV Testing and Surveillance," GL 95-07 "Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves," and GL 96-05 "Periodic Verification of Design-Basis Capability of Safety-Related Power-Operated Valves." The inclusion of the MOVs from the PSW system in the ONS GL 89-10 MOV program provides added assurance that these active MOVs will be capable of performing their safety function under those conditions for which

' the system is designed. On the basis that all MOVs from the PSW system will be tested in accordance with the licensee's 1ST and GL 89-10 MOV programs, the NRC staff determined that the testing of MOVs for PSW is in conformance with the ASME OM Code, and NRC regulatory requirements.

Finally, the licensee stated that there are no new requests to the NRC for alternatives or relief from the ASME OM Code requirements associated with the PSW system modification. Since no relief or alternatives are requested, the NRC staff concludes that the 1ST of PSW pumps and valves will be performed in accordance with ASME OM Code requirements.

The NRC staff also reviewed the proposed new TS section as related to 1ST of the PSW pumps and valves (TS 3.7.10, "Protected Service Water System.") Specifically, in this newTS section, SR 3.7.10.3 is proposed to verify, in accordance with 1ST program, that the developed pead of the PSW pumps (both primary and booster pumps), at the flow test point, is greater than or equal to the required developed head. The NRC staff finds that this proposed TS addition is in conformance with 10 CFR 50.36, and that SR 3.7.10.3 is consistent with the ASME OM Code requirements as well as the proposed revisions to the licensee's 1ST program, and is, therefore, acceptable.

3.2.4 lnservice Testing Conclusion Based on the above described review, the NRC staff finds that the 1ST of certain pumps and valves used for PSW operation meets the requirements of 10 CFR 50.55a(f), the ASME OM Code, GL 89-10, GL 95-07, and GL 96-05. As a result, the NRC staff concludes that the proposed TS changes and UFSAR additions relating to the inservice testing of PSW system components are acceptable for ONS Units 1, 2, and 3.

3.3 Reactor Safety Performance Requirements of the PSW System Within the scope of this license amendment request, the licensee proposes to install a PSW system. The reactor safety performance requirements for the PSW system are to provide 375 gallons per minute protected service water flow to each unit's SGs at a pressure of 1082 psig within 15 minutes following a postulated HELB in the Turbine Building. The PSW system must also provide a means to establish RCP seal cooling within 20 minutes following a HELB in the Turbine Building. These requirements are provided in the draft TS Bases for the PSW system, which the licensee provided in response to Request for Additional lnform?tion (RAI) 107 in to the December 16, 2011, letter (ADAMS Accession N6. ML12003A070).

These requirements are in place to ensure that the PSW provides adequate protection following the postulated design basis HELB in the Turbine Building. In its response to RAI 152, provided in the enclosure to a July 11, 2012, transmittal (ADAMS Accession No. ML12195A325), the licensee stated that the PSW design basis HELBs include the following:

The PSW system is relied upon to mitigate postulated HELBs, where pipe whip or jet impingement interactions can result in the loss of main steam system integrity and the loss of the 4160 VAC Essential Power system located inside the Turbine Building.

Secondly, the PSW system is relied upon to mitigate a postulated Feedwater Line Break inside the Turbine Building, which can result in the loss of all AC power to the affected unit, and a loss of the' Turbine-Driven EFW pump on the affected unit. This scenario results in a complete loss of normal and emergency feedwater necessary for secondary system decay heat removal.

In addition, the PSW system is relied upon to mitigate postulated HELBs that can result in a loss of the low pressure service water system and EFW pumps located in the Turbine Building Basement due to flooding.

The NRC staff's review was performed to determine whether the licensee has demonstrated that the design basis requirements for the PSW system provide defense-in-depth and added assurance of the protection of public health and safety. The NRC staff reviewed the licensee's analyses of the PSW design basis events as discussed in the following sections of this safety evaluation. The review scope under this section (3~3) is limited to the licensee's analysis of immediate reactor effects following the PSW design basis events.

3.3.1 Double-Ended Main Steam Line Break 3.3.1.1 Staff Evaluation As discussed in Chapter 15 of the Oconee UFSAR, the limiting MSLB for the plant licensing basis is a double-ended guillotine rupture. The single MSLB is analyzed both with and without offsite power. In the scenario without offsite power, the analysis is completed at the beginning of cycle which conservatively predicts the departure from nucleate boiling as the reactor coolant pumps coast down. Conversely, the scenario with offsite power is analyzed at the end of cycle to maximize the positive reactivity addition resulting from the RCS cooldown or any return to power.

The acceptance criteria are that the core will remain intact for effective core cooling and that the offsite doses will be within 10% of the 10 CFR Part 100 limits.

The PSW system performance requirements are established to mitigate postulated HELBs. *The HELB scenario is one in which pipe whip or jet impingement can lead to loss of main steam system integrity. For evaluating the adequacy of the PSW system, therefore, the licensee selected a double-ended MSLB. The PSW system is also used when there is a loss of the 4160 VAC essential power system in the Turbine Building. In this analysis; a unit trip is assumed at the initiation of the event. This assumption is conservative when analyzing the effects of overcooling.

The double-ended MSLB with loss of the 4kV essential auxiliary power system, as described in Chapter 7 of ONDS-351, "Oconee Nuclear Station Units 1, 2, & 3, Analysis of Postulated High Energy! Line Breaks (HELBs) Outside of Containment," Revisi~n 2 (transmitted in the licensee's letter dated December 16, 2011 ), is the limiting overcooling event which challenges the PSW system. In this scenario, no credit is taken for main steam isolation valve (MSIV) closure as a means to mitigate the overcooling event. Two conditions of the overcooling event are analyzed, one with the RCPs operating and one without the RCPs operating. The only method of borated water addition is through the core flood tanks (CFTs).

The purpose of the double-ended MSLB analysis is to demonstrate that core protection is maintained prior to and immediately following a reactor trip. After the reactor trip, the double-ended MSLB with loss of offsite power analyses are applied to demonstrate that the

. minimum DNBR remains higher tha'n the safety limit.

The double-ended MSLB is analyzed with an initial power level of 102% full power and a reactor trip at 0.1 seconds. The RCS rapidly cools and injection of borated water from the CFTs occurs around 55 seconds. Actions taken by the operator are via the PSW system to maintain stable RCS temperature and pressure until power can be restored to the low pressure injection (LPI) pump, LPSW pump, and the decay heat drop line isolation valves. The PSW system maintains the stable conditions by supplying power to the HPI pump and HPI valves. This in turn provides reactivity control, RCP seal cooling, and RCS inventory recovery. The PSW can also supply power to the pressurizer heaters to maintain RCS pressure once water level is restored to the pressurizer, maintaining pressure control. The PSW system also provides power to the reactor vessel (RV) Head Vent and RCS Loop High Point vent valves, allowing for reactivity and inventory control.

The licensee used the RETRAN-3D.code to simulate the nuclear steam supply system response to the double-ended MSLB transient and to provide dynamic core conditions to the VIPRE (DPC-NE-3000-PA) thermal-hydraulic code. The licensee further discusses the use of RETRAN-3D and VI PRE in its response to RAI 156, dated July 11, 2012 (ADAMS Accession No. ML12195A325). The difference between the single MSLB RETRAN-3D model and the double-ended MSLB RETRAN-3D model is the break in two steam lines rather than one. VI PRE is used to determine the minimum DNBR.

The licensee's RAI response dated July 11, 2012 (ADAMS Accession No. ML12195A325) describes the input parameters and assumptions used in the double-ended MSLB with loss of the 4kV essential power system analyses. The limiting DNBR case of a double-ended MSLB with loss of offsite power is also discussed. In this case, the initial and boundary conditions are nearly identical to the single MSLB without offsite power discussed in USFAR Section 15.13.3, with the exception being the double-ended MSLB modeling. The figures provided describe the results of these analyses. Included are plots of reactor power, RCS pressure, CFT injection, Tave. break flow and reactivity.

The limiting case of a double-ended MSLB with loss of offsite power results in a minimum DNBR of 1.661, which is above the DNBR safety analysis limit of 1.41. The double-ended MSLB with loss of offsite power is determined to be the limiting DNBR case because the RCPs are immediately lost in this scenario. There is no return to criticality and the RCS pressure drops below 200 psig within the first 150 seconds of the transient with reactor coolant pumps running.

The limiting Oconee MSLB cases demonstrate that the calculated minimum DNBR remains above the DNB safety limit, and there is no return to criticality, thus ensuring that fuel rod failure 1 does not occur.

[

3.3.1.2 Staff Findings The NRC staff reviewed the licensee's analysis of the double-ended MSLB and concludes that the licensee's analysis is performed using acceptable analytical models, and that the results meet the DNB design basis criteria.

Based on these considerations, the NRC staff concludes that the licensee has demonstrated that the PSW design basis requirements are established such that the system will provide acceptable mitigation following HELBs such as a double-ended MSLB.

3.3.2 Feedwater System Pipe Breaks in the Turbine Building 3.3.2.1 Summary of Licensee's Analysis Certain main feedwater line breaks inside the Turbine Building can result in a complete loss of electrical power to the affected unit, as described in the licensee's report, ONDS-351, Revision 2, pages 7-8. This essentially creates a blackout condition on the affected unit. These breaks also result in a loss of the turbine-driven (TO) EFW pump on the affected unit. These breaks can result in ruptures to the CCW and LPSW piping. Ruptures in the CCW and LPSW piping can lead to Turbine Building flooding events, which can fail all EFW pumps and LPSW pumps. The effect is that the unit experiences a complete loss of secondary system decay heat removal. The event leads to an overheating condition for the RCS. The RPS will trip the reactor on the loss of main feedwater pumps or on high RCS pressure. The pressurizer code safety valves are credited to relieve pr~ssure to maintain RCS pressure below the safety limit. The main steam lines are assumed to remain intact with only the main steam atmospheric relief valve~s (MSRVs) lifting when the high main steam pressure setpoin~ is reached to maximize the RCS heatup.

3.3.2.2 Staff Evaluation The discussion in ONDS-351, Revision 2, Chapter 7, did not explain which postulated HELBs had been identified as limiting with,respect to establishing the reactor safety performance requirements for the PSW system. In response to RAI 152 (July 11, 2012; ADAMS Accession No. ML12195A325), the licensee identified this postulated event as the limiting event with respect to RCS overheating, based on the challenging consequential failures identified above. Other postulated HELBs resulting in overheating, which do not involve the failure of the 4160 VAC essential power system, do not impose such a severe challenge with respect to RCS overheating.

Based on the fact that the identified feedwater line break causes the most coincident damage, the NRC staff finds the licensee's identification of the limiting event acceptable.

ONDS-351 does not discuss the FWLB analysis at such a level of detail as to permit staff review.

During an audit, the licensee provided calculation OSC-2310, "SSF Design Bases Evaluation,"

Revision 23, for NRC staff review. This document evaluates the capabilities of the SSF, which provides mitigative capabilities similar to that of the PSW; i.e., it provides 350 gpm auxiliary service water to the SGs and establishes a makeup source for RCP seal cooling. The calculation demonstrates acceptable performance in consideration of the acceptance criteria discussed above. That analysis was docketed in the licensee's response to NRC staff RAI 153, dated July 11, 2012 (ADAMS Accession No. ML12195A325).

The initiating event chosen for the analysis is ~ loss of normal feedwater, since the FWLB is upstream of the feedwater system check valves and no loss of SG secondary inventory would occur following the postulated FWLB. THe FWLB would then cause the loss of feedwater due to its concurrent loss of 4160 VAC power. Although the analyzed event assumes the reactor trips due to an anticipatory loss of main feedwater trip signal, the licensee confirmed in the response to RAI 155 that the difference between the anticipatory trip and the consequential control rod insertion due to loss of 4160 VAC vital power would be insignificant with respect to the challenge on the reactor coolant system pressure boundary integrity 1.

The licensee stated in response to RAI 153, dated July 11, 2012, that the event was modeled using the RETRAN modeling tool, generally in accordance with NRC-approved Duke reload safety analysis methods described in DPC~NE-3000P-A and DPC-3005P-A.

The model assumes that the reactor has been operating for an entire cycle at 102% of the current licensed thermal power leveL All other reactor state parameters are assumed at their nominal values. For design basis accident and transient analysis, the NRC staff does not typically accept the realistic modeling of reactor state conditions unless the associated uncertainty is quantified explicitly; however, the licensee provided supplemental information to justify the chosen modeling approach. In response to RAI 154, dated July 11, 2012, the licensee explained that reactor power, and the subsequent decay heat associated with it, is the dominant initial condition driving the plant response until the time the SSF auxiliary service water (or PSW) is actuated. Differences in initial pressure may result in a time shift on the order of seconds when the pressurizer power-operated relief valve or safety valve will lift or reseat, but will not change the expected long-term response.

Error or uncertainty with other initial conditions is similarly insignificant. The NRC staff accepts the licensee's response, because it clarifies that any uncertainties in applying the realistically modeled reactor state parameters are generally insignificant in the overall context of the modeled FWLB event.

3.3.2.3 Staff Findings The NRC staff has reviewed the licensee's analyses of the FWLB event, and concludes that the licensee's analyses have adequately accounted for operation of the plant with the proposed PSW 1 The analyzed event credits the mitigation of RCS overpressurization using the pressurizer power-operated relief valve. In addition to this, pressurizer safety valves are available to provide RCS overpressure protection.

system modifications and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated that the reactor protection and safety systems will ensure that the reactor coolant system pressure will remain below its safety limit during this event. Based on these considerations, the NRC staff concludes that the licensee's Protected Service Water system will meet its reactor safety performance requirements at Oconee Nuclear Station. The NRC staff therefore finds that the proposed TS requirements and UFSAR supplements pertaining to the proposed PSW system are acceptable.

The licensee's analyses have demonstrated, as discussed above, that the reactor coolant system remains in an acceptable condition following the limiting postulated HELBs until the PSW system can be actuated, natural circulation core cooling can be established with decay heat removal to*

the SGs via cooling from the PSW system, and high pressure injection can be established for RCP seal cooling. Natural circulation core cooling is addressed in Oconee UFSAR Chapter 15, Section 6.7, "Natural Circulation Cap9bility Analysis." As natural circulation flow capabilities *re addressed in this UFSAR section, the NRC staff did not consider natural circulation core cooling as part of its review for the PSW system.

3.4 Electrical Review As discussed in Section 1.0 of this SE, the licensee made chang~s to the ONS AC power system to supply power to the PSW system, and those changes were made under the provisions of 10 CFR 50.59. Therefore, those changes are not within the scope of the staff's review of this amendment request, and are subject to inspection under the NRC's. Reactor Oversight Process.

Similarly, the staff did not review the PSW electrical system interfaces with safety-related onsite AC and DC power systems. The NRC staff reviewed the aging management provisions for the PSW electrical system cables and the design of the direct current (DC) system for the PSW system. The staff's review of the newTS requirements for the PSW system is provided in Section 3.8 of this SE.

The PSW system is not a safety-related system and no credit is taken in the safety analyses for PSW system operation following design basis events. However, the NRC staff notes that the licensee provided the following statements in the proposed UFSAR revisions submitted on December 18, 2013:

"The [PSW] system... is electrically independent from the station electrical distribution system," and, "Failures in the PS\\fl/ system will not cause failures or inadvertent operations in existing plant systems."

Confirmation that the installed configuration of the PSW cabling provides adequate independence and separation from any Class 1 E circuits is subject to NRC inspection.

PSW Cable Aging Management In its response to the NRC staff's questions related to preventing the effects of cable submergence, dated July 11, 2012 (ADAMS Accession No. ML12195A325), the licensee committed to performing periodic condition inspections of the cable trenches, duct banks, manholes and drainage systems associated with the cable routing through the underground path I

between the Keowee Hydroelectric Units (KHUs) and the PSW building. Duke Energy also committed to include those cables in the Oconee Cable Aging Management Program. The' licensee included these commitments in Attachment 2 of its letter dated July 11, 2012. In an electronic message dated April29, 2014 (ADAMS Accession No. ML14157A240), the licensee stated the following:

1) Model Work order 2103098 was created for the periodic inspections per Maintenance Procedure MP/0/B/2002/002, "Protected Service Water (PSW) Underground Cable Duct Bank-Drainage System Inspection." Due date for the first inspection is October 1, 2014.

The Keowee Underground Trench is currently inspected by ONS Maintenance Procedure MP/0/B/2002/001, "Inspection of the Keowee Underground Cable Trench Drainage System."

2) The ONS Insulated Cables and Aging Management Program has been revised to include the PSW cables in the scope of cables to be inspected every ten years, similar to other existing inaccessible medium voltage cables present in the plant (per License Renewal).

The NRC staff noted that the licensee described its program to manage the aging effects of medium voltage cables in UFSAR Section 18.3.14, "Insulated Cables and Connections Aging Management Program." Certain aspects of the licensee's program were similar to those described in the Generic Aging Lessons Learned (GALL) Aging Management Program (AMP),

Revision 1, (AMP XI.E3), although the licensee did not commit to adopt the GALL AMP. The licensee's program applies to inaccessible (e.g., in conauit or airect buriea) meaium-voltage cables within the scope of license renewal that are exposea to significant moisture simultaneously with significant voltage (i.e., energizea more than 25% of the time). Similar to the provision of the GALL AMP, Revision 1, the licensee plannea to excluae the meaium voltage cables associatea with the PSW system moaification from their AMP,.as they will not be energizea more than 25% of the time.

Generic Letter 2007-01, "Inaccessible or Unaergrouna Power Cable Failures that Disable Acciaent Mitigation Systems or Cause Plant Transients," regarding cable failures ana cable submergence inspection finaings at several plants, resultea in the NRC staff revising the AMP associated with medium voltage cables in Revision 2 of the GALL report. The updated GALL AMP now incluaes all power cables greater than or equal to 480 volts, and includes those cables in the AMP, regardless of the percentage of time they are normally energizea.

For the new PSW system modification, the NRC staff was concerned that the environmental conaitions in trenches and manholes for the installea cables may cause a decrease in the aielectric strength of the cable insulation and potentially result in failure of the PSW cables. In response to the NRC staff's concern, by letter datea July 24, 2014, the license.e proposea aaaitional changes to the ONS UFSAR, Chapter 18, "Aging Management Programs ana Activities." Specifically, the licensee will revise UFSAR Section 18.3.14, to incluae PSW 13.8 kV cabling in the AMP, as statea in the following paragraphs:

Frequency - Accessible insulatea cables ana connections incluaing Protectea Service Water (PSW) 13.8 kV cables installea in aaverse, localizea environments will be inspected at least once every 1 0 years. Water collection in manholes containing in-scope, medium-voltage cables will be monitorea at a frequency aaequate to prevent the cables from being exposed to significant moisture. The PSW drainage system of the trenches and manholes containing the PSW 13.8 kV cables shall be periodically inspected annually to detect exposure of these cables to significant moisture. The periodic inspections shall inQiude video imaging of the drainage systems of the trench and manholes. If significant moisture is detected, actions shall be taken to correct this condition.

Inaccessible or direct-buried, medium-voltage cables exposed to significant moisture and significant voltage will be tested at least once every 10 ye_ars. The PSW system inaccessible 13.8 kV insulated power cables from the Keowee Hydroelectric station to the PSW building and from the PSW substation to thePSW building (Fant Line) shall be periodically ele_ctrically tested. The initial PSW 13.8 kV cable testing shall be performed prior to declaring the entire PSW system operable and thereafter at a 6 year frequency.

The electrical tests shall follow the cable condition monitoring methods and testing techniques pnovided in Regulatory Guide 1.218 (April2012).

I Based on this proposed discussion in the UFSAR, the staff finds that the power cables associated with the new PSW system modification will be covered under an acceptable condition monitoring program that is capable of identifying aging effects in the cables.

The PSW System 125 Volt DC System In its letter dated December 16, 2011, the licensee stated that the PSW DC system consists of two (2) QA-1 battery banks each consisting of 60 C&D LCY-391ead-calcium flooded cells and an end of duty cycle cell voltage of 1.81 Vdc. Either bank can meet the PSW DC system design basis duty cycle with up to two (2) cells jumpered out. Only one bank is required to align to the PSW DC System. The other battery bank will be maintained on float charge and will be available to be aligned to the PSW DC system by manual breaker operation. The licensee's calculation demonstrated that for all cases, the required equipment will have adequate DC voltage to perform its design function with the battery at the end of life capacity (80% ). The PSW system batteries are sized with adequate design margins consistent with NRC-endorsed IEEE Std. 485-1997 (reaffirmed in 2003), "IEEE Recommended Practice for Sizing Lead-Acid Batteries for Stationary Applications," and extra capacity to allow for future growth on the PSW DC system. The PSW building has two Battery Rooms. Each Battery Room has separate hydrogen removal, and heating and cooling systems. The PSW battery chargers are sized to recharge each battery in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while supplying the largest combined load demands of the various continuous steady state loads after a battery discharge to the bounding design discharge state associated with PSW events.

In its July 11, 2012, response to the NRC staff's questions related to the PSW system battery equalizing voltage, the licensee confirmed thaUhe PSW battery equalizing voltage will not exceed the voltage ratings of the equipment and components connected to the PSW 125 Vdc system, based on the licensee's practice to perform battery equalization off-line. In addition, in its response dated July 20, 2012, the licensee provided a summary table showing DC equipment ratings and the voltage drop summary. The NRC staff concludes that the calculated equipment voltages are bounded by the minimum and maximum equipment rated voltages and are therefore acceptable.

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I In its letter dated November 2, 2012, the licensee clarified the proposed PSW system TS Surveillance Requirement (SR) 3. 7.1 Oa.2 associated with verification of the battery pilot cell

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voltage. The licensee stated that the battery pilot cell voltage will be maintained <::: 2.07 V, which represents the lowest voltage cells in the battery. The NRC staff finds the proposed TS limit for battery pilot cell voltage acceptable, because it provides reasonable assurance that the batteries will be able to provide sufficient voltage to all required components.

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Based on the proposed revisions to the UFSAR and licensee responses discussed above, the NRC staff finds that the PSW DC system design requirements and the AMP provisions for PSW electrical cables provide reasonable assurance that the system will be capable of performing its design function.

3.5 Mechanical/Structural Integrity of PSW System The scope of review for this SE section is related to the structural integrity of the mechanical portion of thy PSW system.

I 3.5.1 Piping and Pipe Supports According to RG 1.29, SSCs that must be designed to remain functional during a seismic event are designated as Seismic Category I and all other SSCs are designed as non-Seismic Category I.

The licensee's designation for Seismic Category I is Seismic I. The designation for non-Seismic Category I is Seismic II.

The PSW pumps (primary and booster) and the PSW pipe header are located in the Auxiliary Building. As noted in the licensee's letter dated July 20, 2012 (ADAMS Accession No..

ML12207A109), the PSW system piping is ONS Class "F" and is designed in accordance with the Code of Record U.S.A.S. B31.1 of 1967. According to ONS UFSAR Section 3.1, Class "F" piping is required to be seism[cally designed.

In its letter dated March 16, 2012 (ADAMS Accession No. ML12081A126), in response to RAI109, the licensee notes that the seismic analysis for the PSW piping system was performed using dynamic modal analysis techniques utilizing the SUPERPIPE computer code. SUPERPIPE was procured from AREVA (Framatome-ANP) by ONS under its QA-1 program, which according to UFSAR Section 3.1.1.1 meets the requirements of 10 CFR 50, Appendix B. ONS design basis response spectra was used for the dynamic seismic analysis of piping per ONS specification OS-027B.00-00-0002, "Specification for the Seismic Displacements and Response Spectra for the Turbine, Auxiliary, Reactor, and Standby Shutdown Facility Buildings," Revision 8.

In its letter dated December 18, 2013, (ADAMS Accession No. ML13358A042), the licensee stated in proposed UFSAR Section 9.7.3.5.2, "Subsystem Seismic Analysis", that ONS earthquake motion is two directional in accordance with UFSAR Section 3.7.2.5. In the July 20, 2012, letter, the licensee indicated that the PSW SSCs were evaluated using the two-directional earthquake with the absolute sum rule, by which the absolute algebraic sum of the vertical (Y) seismic component and the maximum horizontal seismic component (greater of X or Z) are utilized. Therefore, the PSW SSCs have been analyzed for maximum horizontal component (either X or Z) and the vertical component (Y) for seismic loads applied simultaneously. The licensee also designed piping supported from multiple levels or structures using an envelope of the response spectra for all supporting structures.

The NRC staff reviewed the licensee's piping seismic analysis method for Seismic I piping and pipe supports. The review included modeling procedure, multiple levels of supports, interaction of Seismic II over I and overlap of Seismic I to Seismic II piping. The NRC staff noted that evaluating the PSW SSCs, using the two-directional earthquake with the absolute sum rule, is in accordance with the ONS current UFSAR Section 3.7.2.5. The NRC staff finds the licensee's piping seismic analysis method to be acceptable, since it was performed in accordance with the ONS current licensing and design basis, and the spectrum utilized conservatively envelopes spectra for multiple levels of support.

The NRC staff also reviewed the licensee's piping and pipe support load combinations, presented in the licensee's July 20, 2012, letter, for normal, upset and faulted conditions and found them acceptable, since they are in accordance with the ONS current design basis specifications for load combinations in calculating pipe stresses and reaction loads for pipe support design.

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Additionally, the NRC staff reviewed the licensee's July 20, 2012, letter and found that the maximum pipe stress load capacity ratio (L/C) (calculated over allowable) was 0.85 and the maximum equipment nozzles L/C ratio was 0.75. These L/C ratios are less than the maximum allowable ratio of one and, therefore, are acceptable. The July 20, 2012, letter also shows that valve accelerations are less than the allowable of 2.5 g's specified in the ONS design basis Specification, OS-027B.00-00-0001. The licensee evaluated PSW system pipe supports in accordance with the AISC Manual of Steel Construction, 6th Edition, 1963. The 6th Edition of the AISC Steel Construction Manual is in the ONS current licensing and design basis and, therefore, is acceptable.

3.5.2 Seismic 11/1 In addition to the PSW system piping located in the Auxiliary Building, there are two additional piping systems located at opposite ends of the PSW Building. These are the Eyewash and Firehose piping systems, which along with their associated pipe supports were seismically designed by the licensee. In its letter dated AprilS, 2013 (ADAMS Accession No. ML13123A159),

the licensee identified that in accordance with the ONS Piping Design Criteria, PDC-120, "Non-Seismic Interactions", Seismic 1111 walkdowns were performed for these two systems and for the PSW piping system in the Auxiliary Building to identify interaction of these piping systems with non-seismic SSCs. Where non-seismic SSCs were identified as potentially interacting with the seismically designed piping, the seismically designed piping was re-routed to avoid any interference. When it was not possible to re-route seismically designed piping, the interfering sse was relocated.

The NRC staff finds this method for evaluating seismic and non-seismic piping systems and any potential interactions to be acceptable, since it was performed in accordance with PDC-120 and the evaluation shows that Seismic 11/1 issues have been properly identified and resolved.

3.5.3 Seismic I to Seismic II Overlap In its July 20, 2012 letter, for seismically designed piping which becomes non-seismic beyond a certain piping location (such as with the use of a seismic/non-seismic boundary valve), the licensee showed that the piping analysis considered a piping overlap region in which the seismic analysis requirements continue for a sufficient distance in the non-seismic portion of the system.

The overlap region included, as a minimum, five effective restraints in each of the orthogonal directions. This was in accordance with the ONS Specification, OS-0278.00-00-0001, Section 4.3.4.6, for overlapped analysis between seismic and non-seismic piping. The same overlap requirement exists in the ONS UFSAR Section 3.7.2, "Procedure Used for Modeling", for overlaps used to separate piping analysis problems.

The NRC staff finds this overlap method acceptable, since it is in accordance with the ONS licensing and design basis. The methodology also provides adequate protection of the seismic portion of the piping from loads coming from the non-seismic portion and from potential pipe failure of the non-seismic piping section and it meets the intent of RG 1.29 for the interface between Seismic Category I and non-Seismic Category I SSCs. In addition, it accounts for the effects of one piping model on another when separating seismic analysis piping into two models.

3.5.4 HVAC Ductwork. Components and Supports i

In its letters dated July 20, 2012 and April 5, 2013, the licensee showed that they evaluated the structural adequacy of the PSW-credited HVAC system ductwork and components (including air handling units (AHUs) and fans), ductwork supports and component supports. The evaluation included deadweight and seismic loads combined in accordance with the ONS current design basis for normal, upset and faulted conditions.

For structural design and analysis, the licensee qualified the HVAC system ductwork and ductwork components, such as AHUs and fans, utilizing the ASME AG-1 Code, 2003 edition.

ASME AG-1 is a standard acceptable to the NRC staff. In its AprilS, 2013, letter, the licensee identified that it performed a line-by-line comparison and found that sections of ASME AG-1-1997,.

applicable to the PSW HVAC system design, were not changed by the 2003 version.

For structural design and analysis, the licensee qualified new or existing HVAC supports and their connections to existing structures affected by the PSW project utilizing the 6th Edition of the AISC Manual of Steel Construction, which is in the ONS current licensing and design basis.

The NRC staff reviewed the licensee's methodology and criteria used in the PSW-credited HVAC system structural evaluations and found them acceptable. The licensee used the current design basis in the load combination load cases (i.e. normal, upset and faulted conditions) for both deadweight and seismic loads. The licensee also evaluated the PSW-credited HVAC system ductwork and components, including AHUs and fans, in accordance with the ASME AG-1, 2003 Code, which it reconciled with the ASME AG-1, 1997 Edition. RG 1.52, Revision 3 and RG 1.140, Revision 2, both endorse the use of the ASME AG-1, 1997 Edition. The NRC staff also finds the use of the 6th Edition of the AISC for the qualification of PSW HVAC duct supports to be acceptable, since it is part of the ONS current licensing and design basis.

3.5.5 Vibration Monitoring In its letter dated July 20, 2012, the licensee indicated that vibration data for the PSW pump assemblies were obtained prior to delivery and that the pumps will also be monitored during initial testing. The licensee also stated that the Oconee Piping Analysis Engineering Manual, Instruction #13 for Piping Vibration (PAEM-013) and the Duke Engineering Modification Process, both require vibration monitoring of systems in accordance with ASME Standards and Guides for Operation and Maintenance of Nuclear Power Plants (OM-SG), Part 3.

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The NRC staff noted that the guidance of ASME OM-SG, Part 3, is recommended in NUREG-0800, SRP 3.9.2, "Dynamic Testing and Analysis of Systems, Structures and Components," for 'vibration monitoring of SSCs. The NRC staff finds that the licensee's plan for monitoring vibration of the piping and supports for the PSW system is acceptable since it provides reasonable assurance that vibration levels will be monitored to show that levels remain within acceptable limits of ASME OM-SG, Part 3, during initial testing and operation of the PSW system.

3.5.6 Seismic Adequacy of the PSW Cable Tray System In its letters dated January 20, 2012 (ADAMS Accession No. ML12025A124) and March 16, 2012, the licensee noted that the cable tray systems located in the PSW Building; the Auxiliary Building and the Keowee Hydroelectric Station were installed to support the PSW electrical distribution system. To verify the seismic adequacy o( the cable tray systems, the licensee utilized the Seismic Qualification Utility Group (SQUG)' Generic Implementation Procedure (GIP) Revision 3A (SQUG GIP-3A). SQUG GIP-3A is one of the current licensing basis methods, discussed in UFSAR Section 3.1 0, for verifying the seismic adequacy of electrical equipment, including the seismic adequacy of all cable tray supports.

The NRC staff finds that the design of the PSW cable tray system is seismically adequate, based on the system design being in accordance with SQUG GIP-3A, which is acceptable per the ONS licensing basis, as noted in UFSAR Section 3.1 0.

The NRC staff reviewed the licensee's structural evaluations of the PSW system piping, pipe supports, cable tray supports, HVAC ductwork, HVAC components and HVAC supports. Based on information and analysis described above, the NRC staff concludes that the licensee has adequately addressed the structural integrity of these components and has provided reasonable assurance that these PSW system components are structurally adequate to perform their intended design functions and remain in compliance with the ONS Design Criteria 1, 2 and 40 with respect to structural integrity.

3.5.7 Consideration of PSW Pipe Ruptures By letter dated August 7, 2013, the licensee proposed the following wording to replace Section 3.6.1.3 of the ONS UFSAR:

"3.6.1.3 Protected Service Water (PSW) System The PSW system is designed as a standby system for use under emergency conditions. With the exception of testing of the system, the system is not normally pressurized. Testing of the system is infrequent, typically every quarter. In addition, the duration of the test configuration is short, compared to the total plant (unit) operating time. Due to the combination of the infrequent testing and the short duration of the test, pipe ruptures are not postulated or evaluated for the PSW system."

By letter dated December 16, 2011, in response to RAI 75, the licensee stated that the PSW system is only expected to be used under extreme emergencies where normal and emergency systems located inside the Turbine Building are damaged from postulated HELBs or a tornado. The licensee further stated that since the PSW sy~tem is not expected to be operated in excess of 1% of the total plant operating time (for routine testing), the system has been excluded from the postulation of HELBs and critical cracks. The NRC staff considered that the PSW system is not normally pressurized; it will only be tested for short durations at infrequent intervals; and it will only be called upon to perform its function as a back-up system for certain limited accident scenarios in which the credited safety systems are rendered unavailable. For these reasons, the staff agrees that the effects of postulated pipe ruptures on essential equipment need not be evaluated for the PSW system.

3.6 Seismic Review of PSW Structures The scope of the review included in this safety evaluation is related to the structural design of the PSW building and the associated underground electrical duct banks, as described in the LAR and the licensee's responses to the NRC staff's questions. The NRC staff notes that modifications to the PSW building and the associat~d underground electrical duct banks performed in accordance with 10 CFR 50.59 and not described in the LAR and supporting documents are outside the scope of this review and are subject to NRC inspection.

The NRC staff's assessment of the PSW building and underground electrical duct banks is based on continued conformance with the ONS design and licensing basis requirements, as well as the applicable industry codes and standards and the acceptance criteria and guidance contained in the applicable sections of the SRP and RGs.

3.6.1 Licensee's Basis for Acceptance of the PSW Structures The licensee provided supplemental information addressing the seismic design of the PSW Building in letters dated July 11, 2012, July 20, 2012, August 31, 2012, April 5, 2013, December 18, 2013, and February 14, 2014. That information is summarized in this section.

The PSW Building is a reinforced concrete structure with a combination of reinforced concrete and steel grating operating floor. The roof is supported on four sides by the exterior reinforced concrete walls. The concrete slabs are supported by a series of walls and beams around their perimeter at regular intervals. The steel grating is supported on structural steel floor framing members which span between exterior reinforced concrete walls and interior walls.

The PSW Building foundation is comprised of continuous exterior spread footings. The internal walls are supported on spread footings and the battery room is supported on a structural slab on grade. All foundations are constructed on compacted structural fill with the exception of the east exterior wall footing, which is supported on concrete fill that extends below the adjacent CCW pipes.

The PSW Building is designed to withstand natural phenomena as specified in the ONS UFSAR, Section 3.1.2.

The PSW Building and the underground electrical duct banks have been designed in accordance with the requirements of ACI 349-97, as supplemented by RG 1.142, Revision 2. All applicable load conditions, including dead loads (consisting of the weight of the structure plus all equipment),

live loads, wind loads in accordance with ONS UFSAR Section 3.3.1, seismic loads in accordance with ONS UFSAR Section 3.7.1.1, and tornado wind and missile loads in accordance with RG 1. 76, Revision 1, were considered. In addition, for areas that are subject to roadway loads, the loads associated with two American Association of State Highway and Transportation Officials (AASHTO) HS20 trucks passing simultaneously, and the Reactor Coolant Pump Motor transport trailer loadings, were considered in the design of the underground duct banks.

NRC SRP Section 3.8.4, Revision 2, was published in March 2007. The PSW project was initiated in 2006 utilizing SRP Section 3.8.4, Revision 2 (Draft), which endorses the use of ANSI/AISC N690-1984. The 1984 version of the N690 code was applied to the design of the new PSW steel structures, as supplemented by Appendix F of SRP Section 3.8.4 Revision 2 (Draft).

For the design of PSW-associated underground electrical duct banks, the licensee employed the beam on elastic foundation analysis method. Conservative values for soil springs were used that yielded an upper bound of maximum forces on critical sections. Reinforcement required for critical sections was conservatively continued through the length of the reinforced concrete duct bank. The governing loaq cases for the design of the duct bank were tornado mi~sile load combinations. The soil cover was not included in the analysis to resist vertical mfssile strikes.

The anchorage of components was designed in accordance with ACI 349-01, Appendix B, as supplemented by RG 1.199, Revision 0.

The PSW Building is not designed for an external flood event associated with the postulated failure of upstream dams. The PSW system is credited for the mitigation of HELBs in the Turbine Building that may result in internal flooding. In the event of flooding due to a break in the non-seismic CCW piping located in the Turbine Building, the maximum expected water level within the site boundary is 796.5 feet (Oconee UFSAR Section 9.6.3.1 ). The grade level at the entrance of the PSW Building is at Elevation 797.0 feet. Waterstops are installed at the slab-to-wall construction joints. The PSW Building is watertight up to Elevation 797.0 feet.

The foundation overturning factor of safety and foundation sliding factor of safety are 1.5 for an operating basis earthquake (OBE) and wind, and 1.1 for a safe shutdown earthquake (SSE) and tornado, in accordance with SRP Section 3.8.5, Subsection 11.5. The calculated factors of safety against overturning and sliding for the PSW Building exceed these criteria.

The water table in the area of the PSW Building is located at Elevation 752'-0". The PSW Building foundation is located at Elevation 788'-3". The base of the foundation is 36'-3".above the water table in this area; therefore, flotation is not considered in the design of the PSW Building.

The seismic foundation bearing pressures meet the minimum factor of safety (FOS) of 3.0 for static loading and FOS of 2.25 for static plus Maximum Hypothetical Earthquake (MHE) seismic loading. There is no foundation uplift due to the MHE seismic loading.

The licensee investigated the subsurface materials underlying the PSW Building and performed direct soil boring and geophysical exploration, including Seismic Cone Penetrometer Testing (SCPT) and Refraction Microtremor (ReMi) testing. The licensee also integrated the data from the cross-hole velocity testing performed near the site for the Radwaste Building in 1981. Based on all the available data, the soil condition beneath the PSW Building was well understood for determining key geotechnical parameters including thickness, unit weight and shear wave velocity (Vs). In general, the soil beneath the PSW Building can be divided into two layers based on their origins. The top layer is the fill with an average Vs about 700 feet/second; and the bottom layer is the underlying natural soil layer, which gradually transitions from the weathered soil layer (Vs around 850 feet/second) into relatively unweathered rock. The total thickness of soil above the unweathered rock is about 80 ft.

For the PSW Building structural evaluation, the licensee used the 0.15g peak ground acceleration (PGA) MHE response spectra presented in the ONS UFSAR, Figure 2-55,. The ONS terminology of MHE is equivalent to the SSE and the ONS terminology of design basis earthquake (DBE) is equivalent to OBE. All safety-related structures at the ONS site are founded on rock and the corresponding MHE is anchored at 0.1g PGA, and the corresponding response spectrum is shown in Figure 2-53 of the UFSAR. The only Seismic Class 1 structure founded on overburden is the CT4 Block House and the corresponding MHE was anchored at a PGA value of0.15g, shown in the UFSAR, Figure 2-55. The licensee considers using the 0.15g PGA response spectrum for the PSW Building to be consistent with what was used for the CT 4 Block House, which is the only Class 1 ONS structure founded on the overburden.

I I

For the generation of the PSW Building in-structure response spectra (ISRS), the licensee used the time history record of the North-:.South (N-S), May 1940 El Centro earthquake, normalized to a peak acceleration of 0.15g, as the input ground motion for both the vertical and horizontal excitation, consistent with Section 3. 7.1.2 of the ONS UFSAR. The 5 percent damped response spectra corresponding to the time history input ground motioh envelopes Figure 2-55 of the ONS UFSAR and exceeds the soil MHE response spectrum sigrtificantly at key frequency ranges (between 1-20 hertz (Hz)). Therefore, the seismic input motion used for the PSW Building is comparable to the one used at the site for the existing structures.

For both the PSW Building design and ISRS generation,*the licensee analyzed the structure using a three dimensional (3-D) finite element (FE) model considering two sets of boundary conditions.

The first set of boundary conditions considered a fixed base model where the foundation nodes were constrained against translation and rotation. The second boundary condition considered lumped soil springs (LSS) to take into account soil-structure interaction (SSI) effects. For the LSS models, strain-compatible soil properties were used to calculate the lower bound (LB), best estimate (BE), and upper bound (UB) LSS parameters for the PSW Building response analysis.

The vertical and horizontal soil springs were calculated using the formulation provided in ASCE 4-98. The effect of soil layering was considered in determining the foundation vertical and horizontal elastic springs impedance,functions. For the LSS response analysis, the calculated composite modal damping was limited to 20% for any mode. No material damping f9r soil was considered.

The licensee considered 100 percent of the dead (permanent) loads (e.g., weight of structure and equipment), 25 percent of the floor live (short term) loads (e.g., general live load), and 75 percent of the roof snow load for dynamic analysis of the PSW Building. The equipment mass was lumped at the location of the equipment. The mass of cable trays, HVAC ducts, and piping and their supports were modeled as distributed mass on floors and walls.

The licensee used values of 2% damping for steel elements and 5% damping for reinforced concrete elements for the MHE, consistent with the ONS UFSAR Section 3.7.1.3.

The licensee used uncracked concrete properties for the seismic analysis of the PSW Building, consistent with the current design basis seismic analysis of all ONS safety-related structures.

The licensee performed a series of linear dynamic response-spectrum analyses for the design of the PSW Building using the X-, Y-, and Z-excitations individually. The modal responses for these individual analyses were combined based on the complete quadratic combination (CQC) method, consistent with RG 1.92, Section C.1.1.

For developing the PSW Building ISRS, the licensee developed the nodal accelerations* time histories for the X-, Y-, and Z-excitations individually for 19 selected locations in the PSW Building. The generation of the ISRS complies with RG 1.122 guidance relative to the frequency intervals for ISRS generation and ISRS peak widening. The selected locations included the location of major equipment, centers and corners of operating floor and roof slabs, and centers of exterior wall panels. At each of the 19 selected locations, each of the X-, Y-, and Z-direction excitations yields three (X-, Y-, and Z-directions) response time histories. Unwidened response spectra were developed at each of the 19 selected nodes and combined to generate the X-, Y -,

and Z-directionaiiSRS at each node. The X-, Y-, and Z-directionaiiSRS at each of the 19 nodes were ther widened +/-15 percent. The ISRS for all selected nodes1were then enveloped to develop ISRS for the operating floor, the battery room floor, and the PSW Bu!lding roof and the exterior walls. The ISRS were developed for both the fixed base and the LSS analyses at the 19 selected nodes. Finally, the enveloped ISRS from the fixed base and the LSS analyses were enveloped.

The horizontal ISRS are the envelope of the X-and Z-direction ISRS.

The co-directional responses (maximum element forces and ISRS at selected nodes) from the individual X-, Y-, and Z-direction excitation analysis (using the response spectra method or the time history method) were summed using the absolute sum rule to obtain the summed X-component, Y -component, and Z-component of the design responses (maximum element forces and ISRS at selected nodes) as follows:

Rx = {JRxxl + JRxvl + JRxzl)

Rv = {JRvxl + JRvvl + JRvzl)

Rz = {JRzxl + JRzvl + JRzzl)

Where, Rx, Rv, and Rz are the summed X-, Y-, and Z-components of the design response (maximum element force or unwidened ISRS at a selected node); for example, Rzv is the Z-component of the design response due toY-excitation.

ONS is a two-directional earthquake motion plant according to the UFSAR, Section 3.7.2.5.

Therefore, the PSW SSCs are designed/qualified for the two-directional earthquake using the absolute sum combination, i.e., maximum of the absolute sum of (Rx plus Rv) or (Rz plus Rv).

The horizontal and vertical floor response spectra for OBE were defined as one-half ( 1/2) of the corresponding SSE values.

The licensee used the ST AAD-Pro software for the PSW Building finite element modeling and analysis, including response spectrum analysis and time history analysis.

Duke Energy contracted Sargent & Lundy (S&L) to perform the PSW Building seismic response analysis and the ISRS development as a safety-related scope of work to be performed under the S&L QA program. The S&L QA program complies with 10 CFR Part 50, Appendix Band 10 CFR Part 21 requirements.

The ST AAD-Pro software has been validated in accordance with S&L Standard Operating Procedure (SOP)-0204. SOP-0204 governs all software validation and verification (V&V) at S&L and is the implementing procedure for the S&L NQA-1 1994 compliant Nuclear QA Program. S&L has validated (V&V) STAAD-Pro for development of the ISRS using the time history method of analysis. S&L has also validated (V&V) the STAAD-Pro response spectra method of analysis when modal response combinations are performed using the CQC method.

The underground CCW piping system is located adjacent to the PSW Building. The PSW Building foundation wall has been designed and constructed not to exert any lateral load on the CCW piping system. In addition, the RCP Motor Refurbishment Facility and Radwaste Facility are in the proximity of the PSW Building. The potential adverse interaction of these existing structures with the PSW Building has been evaluated and it has been concluded that failure of these structures would affect neither structural integrity nor operability of the PSW Building during

~nd after a design basis seismic event.

1 The horizontal tornado wind and missile load combinations governed the design of the PSW Building external walls below and above grade. The generated pressure due to horizontal tornado missile impact is determined to be much higher than the dynamic soil pressure and the passive soil pressure.

The PSW Building and the PSW underground duct banks have been evaluated for low and high trajectory turbine missiles according to the requirements in the ONS UFSAR, Section 3.5.1.2 and RG 1.115. The location of the PSW Building meets the criteria for low trajectory missile exclusion.

The total square footage of the PSW Building, including the vestibules, is slightly less than 4,000 square feet, satisfying the criteria for high trajectory missile exclusion based on the total area presenting an insignificant risk. The PSW duct banks vulnerability to both high and low trajectory turbine missiles was evaluated. All of those portions of the PSW duct banks that were not excluded from turbine missile loading consideration, based on target location or size, were shown to be adequate missile barriers.

At the interface of the PSW electrical duct banks with the existing structures, a flexible connection was provided to accommodate the expected displacement under the design loading conditions while still precluding water intrusion.

The maximum short-term settlement for the PSW Building under static loading is less than

0. 75-inch. The long-term settlement of 0.5 to 1-inch occurring within 10 years is expected. The foundation, walls, roof, interior walls and slabs of the PSW Building were constructed months before any commodities were run into the building. Therefore, the initial settlement of the building occurred prior to running any commodities into the building or connecting any drain lines to the storm water (yard drain) system.

For the two locations where the High Pressure Service Water (HPSW) system pipe penetrates the PSW Building east foundation wall, 12-inch diameter pipe sleeves were embedded in the foundation wall and the flexible seal between the 6-inch diameter HPSW pipe and the wall sleeve provides more than enough clearance to allow for long-term settlement. For the Plant Drinking Water (PDW) pipe, a 5-inch diameter sleeve was embedded in the south end of the west exterior foundation wall of the PSW Building to accommodate the 3-inch diameter POW pipe. For cables entering and exiting the PSW Building, sufficient flexibility exists within the conduit run to accommodate any long-term settlement effects.

All non QA-1 (non-safety related) components inside the PSW Building (HPSW and POW piping, conduits, monorail, etc.) are qualified as QA-4 (Seismic Category II) and have been seismically designed to preclude adverse interaction with safety-related systems and components.

The only commodity under the PSW Building foundation footprint is a new storm sewer line that is adequately protected by soil cover and concrete encasement from any adverse effects of the PSW Building foundation loads. There is no interface between the new storm sewer pipe and the PSW Building below grade walls.

The PSW Building will be included in the Duke Energy EDM-410 "Inspection Program for Civil Engineering Structures and Components". By including the PSW Building into the EDM-41 0 program, the PSW Building is incorporated into the ONS aging management program for civil structures.

I Relative to the seismic qualification of electrical and mechanical equipment, the licensee stated that: (1) all QA-1 electrical and mechanical equipment was seismically qualified in accordance with IEEE 344-1975 and IEEE 323-1974, as supplemented by RG 1.100, Revision 1, which meets or exceeds the requirements of the ONS UFSAR; (2) seismic qualification was performed using shake table testing, analysis or a combination of testing and analysis considering five OBE followed by one SSE; (3) for testing, the 10% margin specified in IEEE 323 was included; (4) for floor and wall-mounted electrical enclosures, the applicable in-structure response spectra demand was used for the equipment mounting location; and (5) for components added to the existing safety-related electrical enclosures, the in-cabinet response spectra demand for the comp~nent mounting locations were specified.

The licensee provided several samples of seismic qualification of electrical and mechanical equipment to demonstrate compliance with the requirements of IEEE 344 and IEEE 323, including graphical presentation of test response spectra enveloping the required response spectra.-

To accommodate routing of the electrical conduits to the Unit 3 Seismic Class 1 Auxiliary Building, a new penetration in the exterior south wall of the Unit 3 Auxiliary Building was required. The licensee evaluated the effects of this new penetration on the exterior south wall for all applicable loading conditions to demonstrate that this new penetration does not adversely affect its safety function.

3.6.2 Staff Evaluation The NRC staff reviewed the licensee's amendment request and responses to the RAis regarding the design of the PSW building and associated structures. Based on that review, the NRC staff determined that there is reasonable assurance that the PSW Building and the associated underground electrical duct banks will withstand the ONS design basis seismic and tornado loading conditions without impairment of structural integrity or intended design function. The bases for1these determinations are described in this section.

The PSW Building and the underground electrical duct banks have been designed and constructed in accord,ance with QA Condition 1 requirements of the ONS quality assurance program.

The PSW Building structural evaluation is based on the 0.15g PGA MHE response spectrum presented in the ONS UFSAR, Figure 2-55. This evaluation is consistent with the MHE definition of the ONS UFSAR for structures founded on soil. The time history record of the N-S, May 1940 El Centro earthquake is used as the vertical and horizontal components of the input ground motion to develop horizontal and vertical floor response spectra for the PSW Building consistent with the ONS UFSAR, Section 3. 7.1.2. The actual response spectrum used for the PSW Building significantly exceeds the soil MHE for the site between 1 and 20 Hz. Therefore, the seismic design input motion used for the PSW Building is comparable to the one used for the existing soil-founded Class 1 structure, the CT4 Block House, at the site. The NRC staff notes that the design basis MHE response spectrum would be different from a site-specific response spectrum, developed using updated seismic sources, new ground motion prediction equations, and soil amplification. Accordingly, ca license condition was established to ensure that the seismic adequacy of the PSW Building, the ass:ociated underground electrical duct banks, and the 1

systems and components housed within the PSW Building will be re-evaluated consistent with the re-evaluation of other site structures in response to the Fukushima Near-Term Task Force Recommendations. The license condition requires that the licensee perform a seismic probabilistic risk assessment (SPRA) which includes the PSW system, in accordance with the Electric Power Research Institute (EPRI) Report No. 1025287, "Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPJD) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic," (i.e., the SPID report, November 2012) for the ONS. In the SPRA process, the licensee shall expand the Seismic Equipment List (SEL) to include the PSW system. The licensee proposed this license condition in its letter dated April 11, 2014.

The licensee developed the SSE horizontal and vertical floor response spectra and broadened them by 15% in compliance with RG 1.122. In addition, the envelope of the ISRS, at selected points in the PSW Building walls and floors, for both the fixed base and LSS analyses, was used to develop the horizontal and vertical floor response spectrum. The NRC staff finds the developed JSRS for the PSW Building acceptable for this analysis.

The licensee defined the OBE horizontal and vertical design spectra as one-half of the corresponding SSE horizontal and vertical design spectra. The NRC staff finds this acceptable since it is consistent with the ONS UFSAR.

The licensee performed dynamic response spectrum analyses for the PSW Building to determine the maximum seismic response in the PSW Building reinforced concrete and structural steel elements. The response spectrum analyses are performed in two directions (vertical direction plus N-S or E-W horizontal direction) consistent with the ONS UFSAR, Section 3.7.2.5. The modal responses are combined using the CQC method described in RG 1.92. The NRC staff finds the method of combining modal responses and modeling of mass/weight for dynamic analysis acceptable since they are in accordance with RG 1.92 and SRP 3.7.2 The damping values,used in the seismic analysis of the PSW Building for structural steel framing and reinforced concrete elements are consistent with the ONS UFSAR, Section 3.7.1.3. The NRC staff finds this acceptable. The composite modal damping was limited to 20% in the LSS response analysis and the NRC staff finds this acceptable since it is in accordance with SRP 3.7.2 guidance.

SRP 3. 7.2 includes guidance to account for accidental torsion by including an additional eccentricity of plus or minus 5 percent of the maximum building dimension. The licensee, in its letter dated April 5, 2013, indicated 'that accidental torsion was considered in the analysis of the PSW Building. Therefore, the NRC staff finds this acceptable since it is in accordance with the guidance of SRP 3.7.2.

The tornado wind, differential pressure, and missile criteria considered in the design of the PSW Building are in accordance with RG 1. 76, Revision 1 (March 2007). This is consistent with the ONS UFSAR, and with the current NRC staff guidance.

The design of the reinforced concrete elements of the PSW Building and the underground electrical duct banks is in compliance with ACI 349-97, as supplemented by RG 1.142. The tornado wind and tornado missile load combinations governed the design of the PSW Building external walls below and abov~ grade. The PSW Building and the PSW underground duct banks were evaluated for low and high trajectory turbine missiles according to the requirements in the ONS UFSAR, Section 3.5.1.2, and RG 1.115. The NRC staff finds this acceptable.

The design of the structural steel elements of the PSW Building is in compliance 'with ANSI/AISC N690-1984, as supplemented by Appendix F of SRP 3.8.4 (Draft Revision 2), which is acceptable to the NRC staff.

The safety factors for foundation overturning and sliding satisfy the NRC staff's guidance specified in SRP 3.8.5. The licensee indicated that none of the load conditions resulted in foundation uplift and there is a minimum factor of safety of 3.0 for static loading and a minimum factor of safety of 2.25 for static plus MHE loading against the calculated maximum foundation bearing pressure.

The licensee indicated that the details for piping or electrical commodities penetrating the PSW Building walls are water-tight and provide adequate clearance and flexibility to accommodate the expected settlement and the differential movement during a design basis seismic event. The NRC staff finds that this meets acceptable design standards.

The licensee evaluated the potential adverse interaction with the adjacent non-Category 1 structures and concluded that the PSW Building will not be adversely affected during and after a design basis seismic event. The NRC staff finds this acceptable, as it is consistent with the NRC staff guidance described in SRP Section 3.7.2, Subsection 11.8.

All non-safety related components in the PSW Building (pipes, conduits, monorail, etc.) are seismically designed to preclude adverse interaction with safety-related systems and components. The NRC staff finds this acceptable.

The licensee used IEEE 344-1975, as supplemented by RG 1.100, Revision 1, and IEEE 323-1974 for seismic qualification of QA-1 electrical and mechanical equipment, which meets or exceeds the requirements of the ONS UFSAR. The ISRS for the PSW Building and other existing structures, corresponding to the component mounting locations, are used in testing or analysis used for seismic qualification. The anchorage of equipment has been designed in accordance with the ACI 349-01, as supplemented by RG 1.199. The NRC staff finds the use of IEEE 344-1975, as supplemented by RG 1.100, Revision 1, IEEE 323-1974 and RG 1.199 for seismic qualification and mounting of ONS PSW components acceptable.

The PSW Building will be jncluded into the ONS aging management program. This will provide reasonable assurance that the aging effects will be adequately monitored throughout the period of extended operation.

The computer software used for the structural evaluation of the PSW Building and ISRS development were controlled under the S&L QA program, which complies with 10 CFR 50, Appendix B and 10 CFR Part 21 requirements.

The licensee provided information and demonstrated that: (1) the new penetration in the exterior south wall of the Unit 3 Auxiliary Building has been evaluated in compliance with the ONS design and licensing basis; and (2) the construction of this penetration does not adversely affect the safety function of the exterior south wall of the Unit 3 Auxiliary Building. Therefore, this is acceptable to the NRq staff.

The NRC staff notes that the PSW Building is only designed for flood conditions up to elevation 796.5 feet. Therefore, the PSW Building, and systems and components housed within the PSW Building cannot be credited for mitigation of flood scenarios above elevation 796.5 feet.

3.

6.3 NRC Staff Conclusion

The NRC staff concluded that the seismic input motion of the PSW Building is consistent with the site MHE, and therefore, the seismic design is comparable to that of the existing Class 1 structures founded on soil at the ONS site. However, with the consideration of updated seismic sources, new ground motion prediction equations, as well as current understanding of site response, a license condition is imposed for the ONS. This license condition require,s the licensee to perform a SPRA which includes the PSW system, in accordance with the EPRI Report No. 1025287, "Seismic Evaluation Guidance: SPID for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic," for the ONS. In the SPRA, the licensee shall expand the SEL to include the PSW system.

The NRC staff also concluded that there is reasonable assurance that the PSW Building and the associated underground electrical duct banks will withstand the ONS DBE subject to the License Condition, and tornado loading conditions without impairment of structural integrity or intended design function. The PSW Building is only designed for flood conditions up to elevation 796.5 feet. Therefore, the PSW Building, and systems and components housed within the PSW Building cannot be credited for mitigation of flood scenarios above elevation 796.5 feet.

3.7 Human Factors 3.7.1

Background

The PSW primary and booster pumps are located in the Auxiliary Building. For extended operation, the PSW portable pump is designed to provide a backup supply of water to the PSW system in the event of loss of CCW and subsequent loss of CCW siphon flow. To ensure extended operation of the PSW, an alternate cooling system is being provided. A new chilled water system, the AWC system, will be installed that uses portable chillers and permanently installed piping to selected air handling units for the Auxiliary Building. AWC components. such as a pump, chille.rs, air handling units (AHUs) and their respective controls have been placed in locations outside the Turbine Building that are accessible and can be operated locally following an event in the Turbine Building. The licensee described the use of AHUs as a part of the alternate cooling strategy for maintaining an acc.eptable temperature for the control complex and the Auxiliary Building. This cooling strategy credits manual actions.

3.7.2 Description of Manual Actions.

The licensee stated in a response to an RAI dated June 28, 2013, that air handling units are a part of the proposed alternate cooling strategy for maintaining an acceptable temperature for the control complex and the Auxiliary Building. The following is a general list of manual actions that may be required to support placing AWC in service:

Start an AWC pump Start an AWC chiller I

Close normal supply valves to credited AHUs Open AWC supply and return valves to AHU Align AWC power to AHU and start AHUs The following is a general list of manual actions that may be required to support placing containment cooling in service:

Stage diesel powered portable pump at lake Route hose from discharge of portable pump and connect to LPSW supply line to RBCU Connect hose to RBCU LPSW return line and route to yard drain Close normal LPSW supply header isolation valves Start diesel powered pump Start RBCU The licensee performed an analysis indicating that restoration of cooling to the Unit 1 & 2 Control Room is the most time sensitive task and will need to be completed within an estimated four-hour timeframe following the event.

The licensee also provided. descriptions of the methodology used in the development, validation, and verification associated with time critical actions including procedures and manuals. The licensee also provided the detailed surveillances and evaluations employed to validate time critical actions.

3.7.3 Location and Accessibility of Displays and Controls In its RAI response dated December 18, 2013, the licensee stated the AWC system is located outside the Turbine Building envelope. AWC components such as a pump, chillers, air handling units and their respective controls have been placed in locations outside the Turbine Building that are accessible and can be operated locally following a HELB or fire event in the Turbine Building.

The licensee also stated that portions of the LPSW provide containment cooling, which is fed by diesel engine driven pumps located at the lake and uses piping that is outside and )n the penetration rooms. In order to pump the lake water through the RBCUs, it is necessary to isolate the LPSW headers using existing manual valves located in the Turbine Building. Access to the Turbine Building is expected to be restored. in a time frame to support isolation of the LPSW headers.

The licensee stated that the lighting (fixed and/or portable) in the area will be validated to ensure sufficient lighting is available to perform the manual action.

No harsh environmental conditions are anticipated by the licensee in the AB as the Auxiliary Building/Turbine Building wall is a three-hour fire barrier. The validation process will be used to ensure that any required tools, equipment, or keys required for the action are available and accessible. This process includes consideration of self-contained breathing apparatus and personal protective equipment if required.

3.7.4 Actions to Be Completed Prior to Crediting PSW:

I I

The following table lists the actions identified by the licensee 'to be completed prior to placing the PSW system into service.

Required Action Due Date Identify and develop procedures for all Prior to approval of procedures and the manual actions that will be credited as a part AWC system being placed into service of the alternate cooling strategy Develop training to ensure individuals' Prior to approval of procedures and the ability to perform these tasks in the required AWC system being placed into service times in accordance with Duke Energy Nuclear System Directive (NSD)- 514 of the Nuclear Policy Manual This include classroom and On-The-Job training for credited manual actions Ensure control of time critical actions to Prior to approval of procedures and the assure the required times can be met by AWC system being placed into service individuals performing credited manual actions in accordance with Duke Energy NSD-514 of the Nuclear Policy Manual Validate and verify the procedures Prior to approval of procedures and the developed for the identified manual actions AWC system being placed into service associated with proposed alternate cooling strategy in accordance with NSD 705 of the Nuclear Policy Manual and Operations Management Procedure 4-02 Assess environmental conditions such as Prior to approval of procedures and the radiological hazards, high AWC system being placed into service temperature/humidity, egress paths, and smoke/toxic gases (resulting from fire) will be assessed for the credited manual actions 3.7.5 Human Factors Conclusion Based on the statements provided by Duke Energy, i.e., that the new credited manual actions are

. to be performed within an estimated four hour window of time, and that appropriate administrative controls will be applied to procedures and training to ensure that the, actions will be completed, the NRC staff concludes that the proposed LAR is acceptable with respect to human factors considerations.

3.8 Technical Specifications for the PSW System The licensee proposed the addition of several TSs for the PSW system. The current version supersedes all previous versions and was submitted by letter dated August 7, 2013 (ADAMS ML13228A268). The proposed PSW TSs consist of TS 3.7.1 0, "Protected Service Water (PSW)

System," TS 3.7.10a, "Protected Service Water (PS~) Battery Cell Parameters," and TS 5.5.22, "Protected Service Water System Battery Monitoring and Maintenance Program."

3.8.1 TS 3. 7.1 0, "Protected Service Water (PSW) System" The PSW is composed of the following Structures Systems and Components:

PSW Building and associated support systems.

Conduit duct bank from the Keowee Hydroelectric Station underground cable trench to the PSW Building.

Conduit duct bank and raceway from the PSW Building to the Unit 3 Auxiliary Building (AB).

Electrical power distribution system from breakers at the Keowee Hydroelectric Station and from the 100 kilovolt (kV) PSW substation (supplied from the Central Tie Switchyard) to the PSW Building, and from there to the AB.

PSW booster pump, PSW primary pump, and mechanical piping taking suction from the Unit 2 embedded Condenser Circulating Water (CCW) System to the Emergency Feed Water (EFW) headers supplying cooling water to the respective unit's Steam Generators (SGs) and High Pressure Injection pump motor bearing coolers.

PSW portable pumping system.

The OPERABILITY of the PSW system provides a diverse means to achieve and maintain safe shutdown by providing secondary side DHR, reactor coolant pump seal cooling, primary system inventory control, and RCS boration for reactivity management during certain plant scenarios that disable the 4160 V essential electrical power distribution system. During periods of very low decay heat the PSW system will be used to establish conditions that support the formation of subcooled natural circulation between the core and the SGs; however, natural circulation may not occur if the amount of decay heat available is less than or equal to the amount of heat removed by ambient losses to containment and/or by other means, e.g., letdown of required minimum HPI flow through the Reactor Coolant (RC) vent valves. When these heat removal mechanisms are sufficient to remove core decay heat, they are considered adequate to meet the core cooling function and systems supporting SG decay heat removal, although available, are not necessary for core cooling.

For PSW OPERABILITY, the following are required:

One (1) primary pump, one (1) booster pump, and one (1) portable pump.

A flowpath taking suction from the Unit 2 CCW piping through the PSW pumping system (including recirculation flowpath) and discharging into the secondary side of each SG and the required HPI pump motor bearing cooler.

TS 3.8.3 required number of 125 Volt Direct Current (VDC) Vital Instrumentation and Control (I&C) Battery Chargers. Note: The Standby battery chargers cannot be credited for PSW OPERABILITY because they are not supplied with PSW power.

One (1) of two (2) PSW batteries and the associated battery charger.

PSW Building ventilation system consisting of ductwork, fans, heaters, fire dampers, tornado dampers, motor-operated dampers and associated controls of the Transformer room AND in-service battery room.

A PSW electrical system power path I from the Keowee Hydroelectric Station.

For PSW OPERABILITY, PSW supplied power is also required for the following:

Either the "A" or "B" HPI pump motor.

PSW portable pump (unless self-powered).

HPI valve needed to align the HPI pumps to the Borated Water Storage Tanks (HP-24).

HPI valves that support RCP seal injection and RCS makeup (HP-26, HP-139, and HP-140).

Pressurizer Heaters (150 kilowatts (kW) above pressurizer ambient heat loss).

Reactor Vessel Head Vent Valves (RC-159 and RC-160).

One (1) RCS Loop High Point Vent Pathway (RC-155 and RC-156 or RC-157 and RC-158).

Required 125 V DC Vital I&C Normal Battery Chargers.

For PSW OPERABILITY, the following instrumentation and controls located in each main control room are also required:

Two (2) high flow controllers (PSW-22 and PSW-24 ).

Two (2) low flow controllers (PSW-23 and PSW-25).

Two (2) flow indicators (one per SG).

One (1) SG header isolation valve (PSW-6).

One (1) HPI seal injection flow indicator.

One (1) "A" HPI train flow indication (from Inadequate Core Cooling Monitoring plasma).

The proposed LCO forTS 3.7.10 would require the PSW System to be Operable in MODES 1 and

2. Operability in MODES 3 through 6 would not be required because the contribution of the PSW system to the reduction of overall plant risk during these MODES is not sufficient to warrant operability requirements.

The proposed LCO is modified by a Note indicating that it is not applicable to Unit(s) until startup from a refueling outage after completion of PSW modifications and after all of the PSW system equipment installed has been tested.

The licensee's TS differ from STS in how LCO 3.0.4 is used. The licensee's LCO 3.0.4 states:

I When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall not be made except when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time. This Specification shall not prevent changes in MODES, or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

Exceptions to this Specification are stated in the individual Specifications.

LCO 3.0.4 is only applicable for entry into a MODE or other specified condition in the Applicability in MODES 1, 2, 3, and 4.

I The licensee's TS Bases describe LCO 3.0.4 as follows:

LCO 3.0.4 establishes limitations on changes in MODES or other specified conditions in the Applicability when an LCO is not met. It precludes placing the unit in a MODE or other specified condition stated in that Applicability (e.g., Applicability desired to be entered) when the following exist:

a. Unit conditions are such that the requirements of the LCO would not be met in the Applicability desired to be entered; and

b. Continued noncompliance with the LCO requirements, if the Applicability were entered, would result in the unit being required to exit the Applicability desired to be entered to comply with the Required Actions. Compliance with Required Actions that permit continued operation of the unit for an unlimited period of time in a MODE or other specified condition provides an acceptable level of safety for continued operation. This is without regard to the status of the unit before or after the MODE change. Therefore, in such cases, entry into a MODE or other specified condition in the Applicability may be made in accordance with the provisions of the Required Actions.

The text for STS LCO 3.0.4 states:

When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made:

a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time;

b. After performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate; exceptions to this Specification are stated in the individual Specifications, or

c. When an allowance is stated in the individual value, parameter, or other Specification.

This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

The STS version of LCO 3.0.4 is a conditional allowance to enter the mode of applicability when the LCO is not met. The licensee uses LCO 3.0.4 as a conditional prohibition from entering the mode of applicability when an LCO is not met. While there is no functional difference between treating LCO 3.0.4 as a conditional prohibition or conditional allowance, there is a difference in how and when LCO 3.0.4 exceptions are listed. The proposed Actions table is preceded by a NOTE that states LCO 3.0.4 is not applicable. The exception for LCO 3.0.4 provided in the NOTE would permit entry into MODES 1 or 2 with the PSW system not OPERABLE. The licensee's justification for this exception is that the PSW is not required to support normal operation of the facility or to mitigate a desi@n basis event. The proposed use of the LCO 3.0.4 note is not I

consistent with STS usage, however it is consistent with the current ONS TS due to the minor differences in how LCO 3.0.4 is described and used in other ONS TS.

The proposed Actions table contains four described Conditions when the LCO would not be met, Conditions A through D.

Condition A describes a situation where the PSW is inoperable. In this situation the licensee would be required to restore the PSW system to operable status within 14 days. The proposed 14-day Completion Time (CT) is based on the SSF Auxiliary Service Water system and reactor coolant makeup (RCMU) system being operable and a low probability of scenarios occurring that would require the PSW system during the 14 day period.

Condition 8 describes a situation where both the PSW System and the SSF are inoperable. The licensee would be required to restore the PSW System to operable status within 7 days in this situation. The proposed 7 day CT is based on the diverse heat removal capabilities afforded by other systems, reasonable times for repairs, and the low probability of scenarios occurring that would require the PSW system during this period.

Condition C describes a situation where the required actions of Conditions A or 8 are not met within the respective CTs. In this situation, and when certain contingency measures have been implemented, the licensee would have 30 days from the discovery of the initial inoperability to restore the PSW System to operable status. Condition C is explicitly modified by a NOTE that states "Condition may only be entered when contingency measures have been implemented."

During ONS Unit 2 refueling outages, the Unit 2 CCW intake piping may be dewatered to accommodate CCW maintenance. The PSW system and SSF both share the Unit 2 CCW intake piping as a source of cooling source for secondary side decay heat removal. Dewatering of the Unit 2 CCW intake piping results in unavailability of the PSW system and the SSF suction source and renders them inoperable. The NRC staff requested further information on the potential duration of such unavailability during Unit 2 CCW maintenance. In response to RAI 150 dated July 11, 2012 (ADAMS ML12195A325), the licensee provided data on the duration of CCW maintenance activities and the resulting total SSF unavailability from previous Unit 2 refueling outages dating back to 1990. In all cases, the total unavailability of the SSF did not exceed 14 days. Total unavailability of the PSW system resulting from CCW maintenance will be equal to that of the SSF. Restoration of the PSW system to operable status is not anticipated to challenge the 30 day completion time allowed by proposed Condition C.

The contingency measures identified by the licensee in the proposed ONS TS 3. 7.10 Bases

prohibit non-essential testing or maintenance of risk significant systems. These systems include the 230 kV switchyard, KHUs and associated power paths, EFW system, HPI system, and Elevated Water Storage Tank (EWST). The contingency measures provide additional assurance that key equipment is available. Appropriate actions for the specific equipment are specified in the applicable TS or the Selected Licensee Commitments manual (the licensee's administrative control document). For example, if the 1A HPI pump becomes inoperable before entry or becomes inoperable after entry, only TS LCO 3.5.2 (HPI), Condition A shall be entered for Unit 1 and the appropriate actions taken until the pump is restored.

The inclusion of contingency measures in the PSW system TS mirrors the existing requirement of 10 CFR 50.65(a)(4) to manage risk associated with maintenance activities. As described in the licensee's response to RAI 151 dated July 11, 2012 (ADAMS ML12195A325), the dewatering of the CCW system is typically a<::companied by a variety of protective measures to reduce plant risk.

Based upon the previous recorded unavailability of the SSF, dewatering of the CCW intake piping is expected to result in inoperability of the PSW system for approximately fourteen days. The expected frequency of this maintenance activity is every 24 months, corresponding to the ONS Unit 2 refueling outage schedule. The inclusion of the contingency measures in the proposed ONS TS 3. 7.10 provides additional assurance that they are carried out before dewatering of the CCW intake piping is initiated. Based on the infrequent occurrence of CCW intake piping dewatering and implementation of contingency measures, the NRC staff finds that the proposed ONS TS 3. 7.10 establishes adequate assurance of PSW system operability and addresses plant risk during anticipated periods of inoperability due to maintenance.

Condition D describes a situation where the required actions of Conditions C are not met within the CT. The licensee would be required to be in MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months in this situation. The proposed 12-hour CT is appropriate to reach the required unit conditions from full power conditions in an orderly manner and without challenging plant systems, considering a three unit shutdown may be required.

The NRC staff reviewed the proposed LCO and determined that it meets the requirements of 10 CFR 50.36(c)(2). The NRC staff also determined that the Actions table and associated Conditions, Required Actions and CTs are acceptable because they are similar to Conditions, Required Actions and Completion times for similar equipment in TS and they are generally in alignment with the NUREG-1430 format.

The licensee proposed SRs 3.7.10.1 through 3.7.10.13 to verify operability of the PSW system.

SR 3. 7.1 0.1 would require the licensee to verify the required PSW battery terminal voltage is greater or equal to the minimum established float voltage at a frequency in accordance with the Surveillance Frequency Control Program.

SR 3. 7.1 0.2 would require the licensee to verify the required Keowee Hydroelectric Station power supply can be aligned to and power the PSW electrical system at a frequency in accordance with the Surveillance Frequency Control Program.

SR 3.7.10.3 would require the licensee to verify that the developed head of PSW primary and booster pumps is greater than or equal to the required developed head at the flow test point at a frequency in accordance with the lnservice Testing Program.

SR 3. 7.1 0.4 would require the licensee to verify PSW battery capacity of the required battery is adequate to supply, and maintain in operable status, the required emergency loads for the design duty cycle when subjected to a battery service test at a frequency in accordance with the Surveillance Frequency Control Program.

)

SR 3. 7.1 0.5 would require the licensee to verify the required PSW battery charger can either supply at least 300 amps at greater than or equal to the minimum established float voltage for more than eight hours or recharge the battery to the fully charged state within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while supplying the largest combined demands of the various continuous steady state loads after a battery di~charge to the bounding PSW event discharge state at ~ frequency in accordance with the Surveillance Frequency Control Program.

SR 3. 7.1 0.6 would require the licensee to verify that the PSW switchgear and transfer switches can be aligned and power both the "A" and "B" HPI pump motors at a frequency in accordance with the Surveillance Frequency Control Program.

SR 3. 7.1 0. 7 would require the licensee to perform functional tests of required power transfer switches used for pressurizer heater, PSW control, electrical panels, vitaii&C chargers, and valves at a frequency in accordance with the Surveillance Frequency Control Program.

SR 3.7.10.8 would require the licensee to verify PSW booster pump and valves can provide adequate cooling water flow to HPI puinp motor coolers at a frequency in accordance with the lnservice Testing Program.

SR 3.7.10.9 would require the licenseeto verify the developed head of the PSW portable pump is greater than or equal to the required developed head at the test point at a frequency in accordance with the Surveillance Frequency Control Program.

SR 3. 7.1 0.10 would require the licensee to verify that the required PSW valves are tested in accordance with the lnservice Testing Program at a frequency in accordance with the lnservice

Testing Program.

SR 3. 7.10.11 would require the licensee to perform a CHANNEL CHECK for each required PSW instrument channel at a frequency in accordance with the Surveillance Frequency Control Program.

SR 3.7.10.12 would require the licensee to perform a CHANNEL CALIBRATION for each required PSW instrument channel at a frequency in accordance with the Surveillance Frequency Control Program.

SR 3. 7.10.13 would require the licensee to verify that the cell plates and racks show no visual indication of physical damage or abnormal deterioration that could degrade battery performance of the required PSW battery cells at a frequency in accordance with the Surveillance Frequency Control Program.

The NRC staff reviewed the proposed SRs. The NRC staff noted that the SRs with a proposed frequency of "In Accordance with the Surveillance Frequency Control Program" had previously been assigned proposed frequencies in alignment with frequencies for similar equipment in the licensee submittals of proposed TS dated December 16, 2011 (ML12003A070) and January 20, 2012 (ML12025A124). The licensee had not fully implemented the Surveillance Frequency Control Program in ONS TS 5.5.21 while the PSW TS were being developed and finalized. The NRC staff determined that the use of the Surveillance Frequency Control Program is appropriate given the initial proposed frequencies were in alignment with frequencies for similar equipment in the ONS TS and the STS. The NRC staff determined that the proposed SRs meet the requirements of 10 CFR 50.36(c)(3), are consistent with requirements for similar components and systems, and also are generally in alignment with the STS in NUREG-1430.

3.8.2 TS 3.7.10a. "Protected Service Water (PSW) Battery Cell Parameters" I

I The PSW direct current (DC) system consists of two batteries, two battery chargers, a distribution center and panel boards. Either battery can be aligned to either battery charger. For PSW DC system OPERABILITY, only one battery and one battery charger is required to be aligned to the PSW DC bus. Each PSW battery consists of 60 cells (nominal) and either battery can meet the PSW DC system design basis duty cycle with up to two cells jumpered out. A minimum of 58 of 60 cells are required for a battery to be considered OPERABLE.

The battery cells are of flooded lead acid construction with a nominal specific gravity of 1.215.

This specific gravity corresponds to an open circuit battery voltage of approximately 124 V for 60 cell battery, i.e., cell voltage of 2.07 Volts per cell (Vpc). The open circuit voltage is the voltage maintained when there is no charging or discharging. Once fully charged with its open circuit voltage < 2.07 Vpc, the battery cell will maintain its capacity for 30 days without further charging per manufacturer's instructions. Optimal long-term performance however, is obtained by maintaining a float voltage 2.20 to 2.25 Vpc. This provides adequate overpotential which limits the formation of lead sulfate and self-discharge. The nominal float voltage of 2.22 Vpc corresponds to a total float voltage output of 133.2 V for a 60-cell battery.

The proposed LCO forTS 3.7.10a, "Protected Service Water (PSW) Battery Cell Parameters" states that, "Battery cell parameters for the required PSW battery shall be within limits." The LCO is applicable when the PSW system is required to be OPERABLE. The battery parameters are required solely for the support of the associated PSW electrical power systems; therefore, battery parameter limits would only be required when the PSW DC power source is required to be operable. Therefore operability requirements for LCO 3. 7.1 Oa would be dictated by LCO 3. 7.1 0 as described above in Section 3.2.1 of this evaluation.

The proposed Actions table is preceded by a NOTE that states LCO 3.0.4 is not applicable. The exception for LCO 3.0.4 provided in the NOTE would permit entry into MODES 1 or 2 with the PSW system not OPERABLE. The licensee's justification for this exception is that the PSW is not required to support normal operation of the facility or to mitigate a design basis event. The proposed use of the LCO 3.0.4 note is not consistent with STS usage, however it is consistent with ONS TS due to the minor differences in how LCO 3.0.4 is described and used in other ONS TS.

The proposed Actions table contains five described Conditions when the LCO would not be met, Conditions A through E.

Condition A describes a situation where the requ(red battery has one or more cells with float voltages less than or equal to 2.07 Volts (V). In this situation, the licensee would be required to perform both SR 3. 7.1 0.1 and SR 3. 7.1 Oa.1 within two hours and restore the affected cell voltage to greater than 2.07 V within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Successful performance of the SRs provides assurance that the battery still has sufficient capacity to perform its function and continued operation for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is appropriate. If SR 3.7.10a.1 is failed, the battery must be immediately declared inoperable.

Condition B describes a situation where the required battery has a float current of more than two amps (A), indicating a partial discharge of the battery may have occurred. In this situation the licensee would be required to perform SR 3. 7.1 0.1 within two hours and restore the battery float current to less than or equal to 2A within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The two hour CT for performance of SR 3. 7.1 0.1 is justified by the time it may take to !diagnose the state of the battery. The 12-hour CT to restore the battery float current is justified by the time it would take to restore a battery to a fully charged condition from a state where it had experienced a temporary discharge, yet all cells remained greater than 2.07 V.

Condition C describes a situation where the required battery has one or more cells with electrolyte levels less than the minimum established design limits. In this situation, the licensee would be required to restore the electrolyte level to above the top plates within eight hours and verify there is no evidence of leakage within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and restore the level to greater than or equal to the minimum established design limits within 31 days. The CTs are justified by the time it may take to perform the level restoration, the leakage inspection, and the level restoration to or above the minimum limits.

Condition D describes a situation where the pilot battery cell electrolyte temperature of the required battery is less than the minimum established design limits. In this situation, the licensee would be required to restore the pilot cell temperature to greater than or equal to the minimum established design limits within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The 12-hour CT is justified by the fact that the battery is sized with margin and would still have sufficient capacity to perform the intended function during restoration of electrolyte temperature.

ConditionE describes a situation where either the required actions for Conditions A, B, C or D are

. not met within the respective CTs, or one or r:nore battery cell float voltages are less than or equal to 2.07 V with a float current greater than 2A. In these situations, the licensee would be required to immediately declare the battery inoperable because sufficient capacity to supply the maximum expected load requirement is not assured.

The NRC staff reviewed the proposed LCO and determined that it meets the requirements of 10 CFR 50.36(c)(2), The NRC staff also determined that the Actions table and associated Conditions, Required Actions and CTs are acceptable because they are similar to Conditions, Required Actions and Completion times for similar equipment in TS and they are generally in alignment with the NUREG-1430 format.

The licensee proposed SRs 3.7.10a.1 through 3.7.10a.6 to verify that the PSW system battery cell parameters are within specified limits.

SR 3. 7.1 Oa.1 would require the licensee to verify the battery float current is less than or equal to 2A at a frequency in accordance with the Surveillance Frequency Control Program. This SR is modified by a Note that states the float current requirement is not required to be met when battery terminal voltage is less than the minimum established float voltage of SR 3. 7.1 0.1. When this float voltage is not maintained, the Required Actions of LCO 3:7.1 Oa ACTION A are being taken, which provide the necessary and appropriate verifications of the battery condition. Furthermore, the float current limit of 2A is established based on the nominal float voltage value and is not directly applicable when this voltage is not maintained.

SR 3. 7.1 Oa.2 would require the licensee to verify the battery pilot cell voltage is greater than 2.07 Vat a frequency in accordance with the Surveillance Frequency Control Program.

SR 3. 7.1 Oa.3 would require the licensee to verify the battery connected cell electrolyte level is greater than or equal to minimum establi.shed design limits at a frequency in accordance with th~

Surveillance Frequency Control Program.

SR 3. 7.1 Oa.4 would require the licensee to verify the battery pilot cell temperature is greater than or equal to the minimum established design limits at a frequency in accordance with the Surveillance Frequency Control Program.

SR 3. 7.1 Oa.5 would require the licensee to verify the battery connected cell voltage is greater than 2.07 Vat a frequency in accordance with the Surveillance Frequency Control Program.

SR 3. 7.1 Oa.6 would require the licensee to verify the battery capacity is greater than or equal to 80% of the manufacturer's rating when subject to a performance discharge test or a modified performance discharge test at a frequency in accordance with the Surveillance Frequency Control Program and every 12 months when the battery shows degradation or has reached 85% of the expected life with capacity less than 1 00% of the manufacturer's rating and every 24 months '

when the battery has reached 85% of the expected life with capacity greater than or equal to 100%

of the manufacturer's rating.

The NRC staff reviewed the proposed SRs. The NRC staff noted that the SRs with a proposed frequency of "In Accordance with the Surveillance Frequency Control Program" had previously been assigned proposed frequencies in alignment with frequencies for similar equipment in the licensee submittals of proposed TS dated December 16, 2011 (ML12003A070) and January 20, 2012 (ML12025A124). The licensee had not fully implemented the Surveillance Frequency Control Program in ONS TS 5.5.21 while the PSW TS were being developed and finalized. The NRC staff determined that the use of the Surveillance Frequency Control Program is appropriate given the initial proposed frequencies were in alignment with frequencies for similar equipment in ONS TS and STS. The NRC staff determined that the proposed SRs meet the requirements of 10 CFR 50.36(c)(3), are consistent with requirements for similar components and systems, and also are generally in alignment with NUREG-1430.

3.8.3 TS 5.5.22, "Protected Service Water System Battery Monitoring and Maintenance Program" The licensee proposed adding TS 5.5.22, "Protected Service Water System Battery Monitoring and Maintenance Program" to the TS. The program would be applicable only to the Protected Service Water battery cells and prov.ides for battery restoration and maintenance, based on the recommendation of IEEE Standard 450-1995. "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications." The program would include the following actions, limits and requirements for preventative maintenance, testing and monitoring for the PSW batteries:

1. Actions to restore battery cells with float voltage ::;; 2.13 V;

2. Actions to determine whether the float voltage of the remaining battery cells is

>2.13 V when the float voltage of a battery cell has been found to be ::;; 2.13 V;

3. Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates;

4. Limits on average electrolyte temperature, battery connection resistance, and battery terminal voltage; and

5. A requirement to obtain specific gravity readings of all cells at each discharge test, consistent. with n1anufacturer recommendations.

The NRC staff reviewed the proposed program and determined that it meets the requirements of 10 CFR 50.36(c)(5) and is also generally in alignment with the battery monitoring and maintenance program. in STS 5.5.17 in NUREG-1430.

3.8.4 Table of Contents The ONS TS Table of Contents pages iii and v were revised to reflect the changes described above. In addition, page iv was revised to restore the references toTS Sections 4.4, 5.0 and 5.1, correcting an administrative error in the issuance of Amendment Nos. 385, 387, and 386, dated April 30, 2014 (ADAMS Accession No. ML14106A418).

3.8.5 Technical Specifications Conclusion The NRC staff evaluated the proposed LCOs, Conditions, Required Actions, Completion Times, SRs and TS Programs. The NRC staff determined that the changes are more restrictive than current requirements. The NRC staff also determined that the format and content of the proposed PSW TS are generally in alignment with NUREG-1430. Finally, the NRC staff determined that the proposed TS comply with 10 CFR 50.36. Therefore, the proposed TS are acceptable.

The regulation at 10 CFR 50.36 states: "A summary statement of the bases or reasons for such specifications shall also be included in the application, but shall not become part of the technical specifications." The licensee may make changes to the TS Bases without prior NRC staff review and approval in accordance with the TS Bases Control Program, TS 5.5.15. Accordingly, along with the proposed TS changes, the licensee submitted TS Bases changes corresponding to the proposed TS changes. The NRC staff determined that TS Bases changes are consistent with the proposed TS changes and provide the purpose for each requirement in the specification consistent with the Commission's Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 2, 1993 (58 FR 39132).

3.9 NRC Staff Summary Conclusion The PSW system is designed as a standby system for use under emergency conditions and provides added "defense in-depth" protection by serving as a backup to existing safety systems.

The PSW system is provided as an alternate means to achieve and maintain safe shutdown conditions for one, two or three units at ONS following postulated scenarios that damage essential systems and components normally used for safe shutdown. On the bases discussed in this SE, and subject to the License Condition and TS requirements imposed by this amendment, the NRC staff finds the proposed amendment acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the South Carolina State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding, which was published in the Federal Register on July 10, 2012 (77 FR 40652). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors:

A. Tsirigotis E. Davidson S.Som M. Hamm J. Huang Date: August 13, 2014 F. Farzam E. Smith P. Sahay B. Parks Y.Li A. Sallman K. Martin D. Woodyatt

ML14206A790 OFFICE NRRILPL2-1/PM NRR/LPL2-1/LA NAME RHall SFigueroa DATE 07/31/14 07/31/14 OFFICE DSS/SRXB/BC* DSS/STSB/BC*

NAME CJackson A Elliott DATE 10/31/12 04/15/14 LPLII-1 R/F RidsNrrDssStsb Resource RidsNrrDoriDpr Resource RidsNrrDeEeeb Resource RidsNrrLASFigueroa Resource YU,NRR ESmith, NRR DWoodyatt, NRR PSahay, NRR RMathew, NRR DSS/SBPB/BC*

DE/EMCB/BC*

GCasto AMcMurtray 04/04/14 04/30/14, 05/02/14 DE/EEEB/BC OGC NLO JZimmerman BHarris 07/31/14 08/11/14

v t e Letter Sequence Approval
TAC:ME7737, (Approved, Closed) TAC:ME7738, (Approved, Closed) TAC:ME7739, (Approved, Closed) TAC:ME7746, (Approved, Closed) TAC:ME7747, (Approved, Closed) TAC:ME7748, (Approved, Closed)
Initiation Request, Request, Request Acceptance... Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement Administration Meeting, Meeting Questions 16 February 2012 11 June 2012 20 December 2012 28 January 2014 Answers 16 December 2011 24 July 2014 Results Approval Other:
MONTHYEAR2013-04-05 [Table View]

ML14206A790

Text

SUBJECT:

Dear Mr. Batson:

Enclosures:

Attachment:

Attachment:

Attachment:

Attachment:

Attachment:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

Background

6.3 NRC Staff Conclusion

Background

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

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