ML17073A067

Issuance of Amendments Revising the Technical Specification Requirements for the Inservice Testing Program (CAC Nos. MF8238 - MF8256)

ML17073A067
Person / Time
Site:	Calvert Cliffs, Dresden, Peach Bottom, Nine Mile Point, Byron, Braidwood, Limerick, Ginna, Clinton, Quad Cities, LaSalle
Issue date:	05/26/2017
From:	Blake Purnell Plant Licensing Branch III
To:	Bryan Hanson Exelon Generation Co
Purnell B, NRR/DORL/LPLIII, 415-1380
References
CAC MF8238, CAC MF8239, CAC MF8240, CAC MF8241, CAC MF8242, CAC MF8243, CAC MF8244, CAC MF8245, CAC MF8246, CAC MF8247, CAC MF8248, CAC MF8249, CAC MF8250, CAC MF8251, CAC MF8252, CAC MF8253, CAC MF8254, CAC MF8255, CAC MF8256
Download: ML17073A067 (229)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 26, 2017 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer (CNO)

Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

BRAIDWOOD STATION, UNITS 1 AND 2; BYRON STATION, UNIT NOS. 1 AND 2; CALVERT CLIFFS NUCLEAR POWER PLANT, UNITS 1 AND 2; CLINTON POWER STATION, UNIT NO. 1; DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3; LASALLE COUNTY STATION, UNITS 1 AND 2; NINE MILE POINT NUCLEAR STATION, UNITS 1 AND 2; PEACH BOTTOM ATOMIC POWER STATION, UNITS 2 AND 3; QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2; R. E. GINNA NUCLEAR POWER PLANT; AND THREE MILE ISLAND NUCLEAR STATION, UNIT 1 - ISSUANCE OF AMENDMENTS REVISING THE TECHNICAL SPECIFICATION REQUIREMENTS FOR THE INSERVICE TESTING PROGRAM (CAC NOS.

MF8238-MF8256)

Dear Mr. Hanson:

The U.S. Nuclear Regulatory Commission (NRC) has issued the following enclosed amendments in response to the Exelon Generation Company, LLC application dated July 26, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16209A218), as supplemented by letter dated October 6, 2016 (ADAMS Accession No. ML16280A402).

1. Amendment No. 191 to Renewed Facility Operating License No. NPF-72 and Amendment No. 192 to Renewed Facility Operating License No. NPF-77 for the Braidwood Station, Units 1 and 2, respectively,

2. Amendment No. 197 to Renewed Facility Operating License No. NPF-37 and Amendment No. 197 to Renewed Facility Operating License No. NPF-66 for the Byron Station, Unit Nos. 1 and 2, respectively,

3. Amendment No. 320 to Renewed Facility Operating License No. DPR-53 and Amendment No. 298 to Renewed Facility Operating License No. DPR-69 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, respectively,

4. Amendment No. 212 to Facility Operating License No. NPF-62 for the Clinton Power Station, Unit No. 1,

5. Amendment No. 254 to Renewed Facility Operating License No. DPR-19 and Amendment No. 247 to Renewed Facility Operating License No. DPR-25 for Dresden Nuclear Power Station, Units 2 and 3, respectively,

6. Amendment No. 223 to Renewed Facility Operating License No. NPF-11 and Amendment No. 209 to Renewed Facility Operating License No. NPF-18 for the LaSalle County Station, Units 1 and 2, respectively,

B. Hanson 7. Amendment No. 227 to Renewed Facility Operating License No. DPR-63 and Amendment No. 161 to Renewed Facility Operating License No. NPF-69 for the Nine Mile Point Nuclear Station, Units 1 and 2, respectively,

8. Amendment No. 313 to Renewed Facility Operating License No. DPR-44 and Amendment No. 317 to Renewed Facility Operating License No. DPR-56 for the Peach Bottom Atomic Power Station, Units 2 and 3, respectively,

9. Amendment No. 266 to Renewed Facility Operating License No. DPR-29 and Amendment No. 261 to Renewed Facility Operating License No. DPR-30 for the Quad Cities Nuclear Power Station, Units 1 and 2, respectively,

10. Amendment No. 124 to Renewed Facility Operating License No. DPR-18 for the R. E.

Ginna Nuclear Power Plant,

11. Amendment No. 290 to Renewed Facility Operating License No. DPR-50 for the Three Mile Island Nuclear Station, Unit 1.

The amendments revise the technical specification requirements for the inservice testing program for each of these facilities.

B. Hanson A copy of the NRC staff's Safety Evaluations are also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, lJ// flvt/1 Blake Purnell, Project Manager Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-456, STN 50-457, STN 50-454, STN 50-455, 50-317, 50-318, 50-461, 50-237, 50-249, 50-373, 50-37 4, 50-220, 50-410, 50-277, 50-278, 50-254, 50-265, 50-244, and 50-289

Enclosures:

1. Amendment No. 191 to NPF-72

2. Amendment No. 192 to NPF-77

3. Amendment No. 197 to NPF-37

4. Amendment No. 197 to NPF-66

5. Amendment No. 320 to DPR-53

6. Amendment No. 298 to DPR-69

7. Amendment No. 212 to NPF-62

8. Amendment No. 254 to DPR-19

9. Amendment No. 247 to DPR-25

10. Amendment No. 223 to NPF-11

11. Amendment No. 209 to NPF-18

12. Amendment No. 227 to DPR-63

13. Amendment No. 161 to NPF-69

14. Amendment No. 313 to DPR-44

15. Amendment No. 317 to DPR-56

16. Amendment No. 266 to DPR-29

17. Amendment No. 261 to DPR-30

18. Amendment No. 124 to DPR-18

19. Amendment No. 290 to DPR-50

20. Safety Evaluation for Braidwood Station, Units 1 and 2; Byron Station, Unit Nos. 1 and 2; Calvert Cliffs Nuclear Power Plant, Units 1 and 2; Clinton Power Station, Unit No. 1; Dresden Nuclear Power Station, Units 2 and 3; LaSalle County Station, Units 1 and 2; Nine Mile Point Nuclear Station, Unit 2; Peach Bottom Atomic Power Station, Units 2 and 3; Quad Cities Nuclear Power Station, Units 1 and 2; and R. E. Ginna Nuclear Power Plant

21. Safety Evaluation for Nine Mile Point Nuclear Station Unit No. 1

22. Safety Evaluation for Three Mile Island Nuclear Station, Unit 1 cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-456 BRAIDWOOD STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 191 Renewed License No. NPF-72

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated July 26, 2016, as supplemented by letter dated October 6, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations 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; 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 amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-72 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 191 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Enclosure 1

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION David J. Wrona, Branch Chief Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: May 26, 2017

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-457 BRAIDWOOD STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 192 Renewed License No. NPF-77

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated July 26, 2016, as supplemented by letter dated October 6, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations 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; 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 amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-77 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 192 and the Environmental Protection Plan contained in Appendix B, both of which are attached to Renewed License No. NPF-72, dated January 27, 2016, are hereby incorporated into the renewed license. The Enclosure 2

licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION c)J ~- if~

David J. Wrona, Branch Chief Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: May 2 6, 2O1 7

ATTACHMENT TO LICENSE AMENDMENT NOS. 191AND192 BRAIDWOOD STATION. UNITS 1 AND 2 RENEWED FACILITY OPERATING LICENSE NOS. NPF-72 AND NPF-77 DOCKET NOS. STN 50-456 AND STN 50-457 Replace the following pages of the Renewed Facility Operating Licenses and Appendix A, Technical Specifications (TSs), with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert License NPF-72 License NPF-72 Page 3 Page 3 License NPF-77 License NPF-77 Page 3 Page 3 TSs TSs 1.1-4 1.1-4 3.4.10-2 3.4.10-2 3.4.14-3 3.4.14-3 3.5.2-4 3.5.2-4 3.6.3-6 3.6.3-6 3.6.6-3 3.6.6-3 3.7.1-2 3.7.1-2 3.7.2-3 3.7.2-3 3.7.5-2 3.7.5-2 5.5-6 5.5-6

(2) Exelon Generation Company, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3) Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels not in excess of 3645 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 191 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Renewed License No. NPF-72 Amendment 191

(2) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. The renewed license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels not in excess of 3645 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 192 and the Environmental Protection Plan contained in Appendix B, both of which are attached to Renewed License No. NPF-72, dated January 27, 2016, are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Renewed License No. NPF-77 Amendment 192

Definitions 1.1 1.1 Definitions INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except Reactor Coolant pump CRCP) seal water injection or leakoff),

that is captured and conducted to collection systems or a sump or collecting tank;

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or

3. Reactor Coolant System CRCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);

b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff} that is not identified LEAKAGE;

c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

MASTER RELAY TEST A MASTER RELAY TEST shall consist of energizing each master relay and verifying the OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

BRAIDWOOD - UNITS 1 &2 1.1 - 4 Amendment 191/192

Pressurizer Safety Valves 3.4.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify each pressurizer safety valve is In accordance OPERABLE in accordance with the INSERVICE with the TESTING PROGRAM. Following testing, lift INSERVICE settings shall be within +/- 1%. TESTING PROGRAM BRAIDWOOD - UN ITS 1 & 2 3.4.10 - 2 Amendment 191/192

RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 -------------------NOTES-------------------

1. Only required to be performed in MODES 1 and 2.

2. RCS PIVs actuated during the performance of this Surveillance are not required to be tested more than once if a repetitive testing loop cannot be avoided.

3. Not required to be performed for RH8701A and B and RH8702A and B on the Frequency required following valve actuation or flow through the valve.

Verify leakage from each RCS PIV is In accordance equivalent to 5 0.5 gpm per nominal inch of with the valve size up to a maximum of 5 gpm at an INSERVICE RCS pressure~ 2215 psig and 5 2255 psig. TESTING PROGRAM, and in accordance with the Surveillance Frequency Control Program ANO Prior to entering MOOE 2 whenever the unit has been in MODE 5 for

~ 7 days, if 1eakage testing has not been performed once within the previous 9 months (continued)

BRAIDWOOD - UNITS 1 &2 3.4.14 - 3 Amendment 191/192

ECCS-Operating 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

SU RV EI LLANCE FREQUENCY SR 3.5.2.3 Verify ECCS locations susce~tible to gas In accordance accumulation are sufficient y filled with with the ivater. Surveillance Frequency Control Program SR 3.5.2.4 Verify each ECCS pump's developed head at In accordance the test flow point is greater than or with the equal to the required developed head. INSERVICE TESTING PROGRAM SR 3.5.2.5 Verify each ECCS automatic valve in the In accordance flow path that is not locked, sealed, or with the otherwise secured in position, actuates to Surveillance the correct position on an actual or Frequency simulated actuation signal. Control Program SR 3.5.2.6 Verify each ECCS pum~ starts automatically In accordance on an actual or simu ated actuation signal. with the Surveillance Frequency Control Program SR 3.5.2.7 Verify, for each ECCS throttle valve listed In accordance below, each position stop is in the correct with the position: Surveillance Frequency Valve Number Valve Function Contro 1 Program SI8810 A,B,C,D Centrifugal Charging System SI8816 A,B,C,D SI System (Hot Leg)

SI8822 A,B,C,D SI System (Cold Leg)

SR 3.5.2.8 Verify, by visual inspection, each ECCS In accordance train containment sump suction inlet is not with the restricted by debris and the suction inlet Surveillance screens show no evidence of structural Frequency distress or abnormal corrosion. Control Program BRAIDWOOD - UNITS 1 &2 3.5.2 - 4 Amendment 191/192

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SU RV EI LLANCE FREQUENCY SR 3.6.3.4 -------------------NOTE--------------------

Valves and blind flanges in high radiation areas may be verified by use of administrative means.

Verify each containment isolation manual Prior to valve, remote manual valve, and blind entering MODE 4 flange that is located inside containment from MODE 5 if and not locked, sealed, or otherwise not performed secured and required to be closed during within the accident conditions is closed, except for previous containment isolation valves that are open 92 days under administrative controls.

SR 3.6.3.5 Verify the isolation time of each automatic In accordance containment isolation valve is within with the limits. INSERVICE TESTING PROGRAM SR 3.6.3.6 Perform leakage rate testing for 8 inch In accordance containment purge valves with resilient with the seals. Surveillance Frequency Control Program SR 3.6.3.7 Perform leakage rate testing for 48 inch In accordance containment purge valves with resilient with the seals. Surveillance Frequency Control Program SR 3.6.3.8 Verify each automatic containment isolation In accordance valve that is not locked, sealed or with the otherwise secured in position, actuates to Surveillance the isolation position on an actual or Frequency simulated actuation signal. Control Program BRAIDWOOD - UNITS 1 &2 3.6.3 - 6 Amendment 191/192

Containment Spray and Cooling Systems 3.6.6 SURVEILLANCE REQUIREMENTS (continued)

SU RV EI LLANCE FREQUENCY SR 3.6.6.4 Verify each containment spray pump's In accordance developed head at the flow test point is with the greater than or equal to the required INSERVICE developed head. TESTING PROGRAM SR 3.6.6.5 Verify each automatic containment s~ray In accordance valve in the flow path that is not ocked, with the sealed, or otherwise secured in position, Surveillance actuates to the correct position on an Frequency actual or simulated actuation signal. Control Program SR 3.6.6.6 Verify each containment s~ray pump starts In accordance automatically on an actua or simulated with the actuation signal. Surveillance Frequency Control Program SR 3.6.6.7 Verify each containment cooling train In accordance starts automatically on an actual or with the simulated actuation signal. Surveillance Frequency Control Program SR 3.6.6.8 Verify each spray nozzle is unobstructed. Foll owing maintenance that could result in nozzle blockage OR Following fluid fl ow through the nozzles SR 3.6.6.9 Verify containment spray locations In accordance susceptible to gas accumulation are with the sufficiently filled with water. Surveillance Frequency Contra 1 Program BRAIDWOOD - UNITS 1 &2 3.6.6 - 3 Amendment 191/192

MSSVs 3.7.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.l Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND OR B.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> One or more steam generators with ~ 4 MSSVs inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3. 7 .1.1 -------------------NOTE--------------------

Only required to be performed in MODES 1 and 2.

Verify each required MSSV lift setpoint per In accordance Table 3.7.1-2 in accordance with the with the INSERVICE TESTING PROGRAM. Following INSERVICE testing, lift setting shall be within+/- 1%. TESTING PROGRAM BRAIDWOOD - UNITS 1 &2 3.7.1-2 Amendment 191/192

MS IVs 3.7.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 -------------------NOTE--------------------

Only required to be performed in MODES 1 and 2.

Verify closure time of each MSIV is In accordance

~ 5 seconds. with the INSERVICE TESTING PROGRAM SR 3. 7. 2. 2 . - - - - - - - - - - - -- - - - -- -NOTE- --------------- ----

Only required to be performed in MODES 1 and 2.

Verify each actuator train actuates the In accordance MSIV to the isolation position on an actual with the or simulated actuation signal. Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.7.2 - 3 Amendment 191/192

AF System 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 Verify each AF manual, power operated, and In accordance automatic valve in each water flow path, with the that is not locked, sealed, or otherwise Surveillance secured in position, is in the correct Frequency position. Control Program SR 3.7.5.2 Verify day tank contains ~ 420 gal of fuel In accordance oi 1. with the Surveillance Frequency Control Program SR 3.7.5.3 Operate the diesel driven AF pump for In accordance

~ 15 minutes. with the Surveillance Frequency Control Program SR 3.7.5.4 Verify the developed head of each AF pump In accordance at the flow test point is greater than or with the equal to the required developed head. INSERVICE TESTING PROGRAM SR 3.7.5.5 Verify each AF automatic valve that is not In accordance locked, sealed, or otherwise secured in with the position, actuates to the correct position Surveillance on an actual or simulated actuation signal. Frequency Control Program SR 3.7.5.6 Verify each AF pump starts automatically on In accordance an actual or simulated actuation signal. with the Surveillance Frequency Control Program (continued)

BRAIDWOOD - UNITS 1 &2 3.7.5 - 2 Amendment 191/192

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 DELETED BRAIDWOOD - UNITS 1 &2 5.5 - 6 Amendment 19l/l9 2

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-454 BYRON STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 197 Renewed License No. NPF-37

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated July 26, 2016, as supplemented by letter dated October 6, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations 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; 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 amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-37 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 197 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Enclosure 3

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULA TORY COMMISSION David J. Wrona, Branch Chief Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: May 26, 201 7

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-455 BYRON STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 197 Renewed License No. NPF-66

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated July 26, 2016, as supplemented by letter dated October 6, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations 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; 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 amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-66 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A (NUREG-1113), as revised through Amendment No. 197, and the Environmental Protection Plan contained in Appendix B, both of which were attached to Renewed License No. NPF-37, dated November 19, 2015, are hereby incorporated into this Enclosure 4

renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

~i~ro2 Br::~

Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: May 26, 201 7

ATTACHMENT TO LICENSE AMENDMENT NOS. 197 AND 197 BYRON STATION. UNIT NOS. 1 AND 2 RENEWED FACILITY OPERATING LICENSE NOS. NPF-37 AND NPF-66 DOCKET NOS. STN 50-454 AND STN 50-455 Replace the following pages of the Renewed Facility Operating Licenses and Appendix A, Technical Specifications (TSs), with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove License NPF-37 License NPF-37 Page 3 Page 3 License NPF-66 License NPF-66 Page 3 Page 3 TSs TSs 1.1-4 1.1-4 3.4.10-2 3.4.10-2 3.4.14-3 3.4.14-3 3.5.2-4 3.5.2-4 3.6.3-6 3.6.3-6 3.6.6-3 3.6.6-3 3.7.1-2 3.7.1-2 3.7.2-3 3.7.2-3 3.7.5-2 3.7.5-2 5.5-6 5.5-6

(2) Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Updated Final Safety Analysis Report, as supplemented and amended; (3) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. The renewed license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels not in excess of 3645 megawatts thermal (100 percent power) in accordance with the conditions specified herein.

(2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 197 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Deleted.

(4) Deleted.

Renewed License No. NPF-37 Amendment No. 197

(2) Pursuant to the Act and 10 CFR Part 70, to receive; possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Updated Final Safety Analysis Report, as supplemented and amended; (3) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. The renewed license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions*specified or incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels not in excess of 3645 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2) Technical Specifications The Technical Specifications contained in Appendix A (NUREG-1113),

as revised through Amendment No. 197, and the Environmental Protection Plan contained in Appendix B, both of which were attached to Renewed License No. NPF-37, dated November 19, 2015, are hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Renewed License No. NPF-66 Amendment 197

Definitions 1.1 1.1 Definitions INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except Reactor Coolant pump (RCP) seal water injection or leakoff),

that is captured and conducted to collection systems or a sump or collecting tank;

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with .the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or

3. Reactor Coolant System CRCS) LEAKAGE through a Steam Generator to the Secondary System (primary to secondary LEAKAGE);

b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE;

c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

MASTER RELAY TEST A MASTER RELAY TEST shall consist of energizing each master relay and verifying the OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

BYRON - UNITS 1 &2 1.1 - 4 Amendment 197/197

Pressurizer Safety Valves 3.4.10 SURVEILLANCE REQUIREMENTS SU RV EI LLANCE FREQUENCY SR 3.4.10.1 Verify each pressurizer safety valve is In accordance OPERABLE in accordance with the INSERVICE with the TESTING PROGRAM. Following testing, lift INSERVICE settings shall be within+/- 1%. TESTING PROGRAM BYRON - UNITS 1 &2 3.4.10 - 2 Amendment 197 /197

RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 -------------------NOTES-------------------

1. Only required to be performed in MODES 1 and 2.

2. RCS PIVs actuated during the performance of this Surveillance are not required to be tested more than once if a repetitive testing loop cannot be avoided.

3. Not required to be performed for RH8701A and B and RH8702A and B on the Frequency required following valve actuation or flow through the valve.

Verify leakage from each RCS PIV is In accordance equivalent to s 0.5 gpm per nominal inch of with the valve size up to a maximum of 5 gpm at an INSERVICE RCS pressure~ 2215 psig ands 2255 psig. TESTING PROGRAM, and in accordance with the Surveillance Frequency Contra 1 Program Prior to entering MODE 2 whenever the unit has been in MODE 5 for

~ 7 days, if leakage testing has not been performed once within the previous 9 months (continued)

BYRON UNITS 1 & 2 3.4.14-3 Amendment 197/197

ECCS-Operating 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2.3 Verify ECCS locations susce~tible to gas In accordance accumulation are sufficient y filled with with the water. Surveillance Frequency Control Program SR 3.5.2.4 Verify each ECCS pump's developed head at In accordance the test flow point is greater than or with the equal to the required developed head. INSERVICE TESTING PROGRAM SR 3.5.2.5 Verify each ECCS automatic valve in the In accordance flow path that is not locked, sealed, or with the otherwise secured in position, actuates to Surveillance the correct position on an actual or Frequency simulated actuation signal. Control Program SR 3.5.2.6 Verify each ECCS pum~ starts automatically In accordance on an actual or simu ated actuation signal. with the Surveillance Frequency Control Program SR 3.5.2.7 Verify, for each ECCS throttle valve listed In accordance below, each position stop is in the correct with the position: Surveillance Frequency Valve Number Valve Function Control Program SI8810 A,B,C,D Centrifugal Charging System SI8816 A,B,C,D SI System (Hot Leg)

SI8822 A, B,C ,0 SI System (Cold Leg)

SR 3.5.2.8 Verify, by visual inspection, each ECCS In accordance train containment sump suction inlet is not with the restricted by debris and the suction inlet Surveillance screens show no evidence of structural Frequency distress or abnormal corrosion. Control Program BYRON - UNITS 1 &2 3.5.2 - 4 Amendment 197/197

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SU RV EI LLANCE FREQUENCY SR 3.6.3.4 -------------------NOTE--------------------

Valves and blind flanges in high radiation areas may be verified by use of administrative means.

Verify each containment isolation manual Prior to valve, remote manual valve, and blind entering MODE 4 flange that is located inside containment from MODE 5 if and not locked, sealed, or otherwise not performed secured and required to be closed during within the accident conditions is closed, except for previous containment isolation valves that are open 92 days under administrative controls.

SR 3.6.3.5 Verify the isolation time of each automatic In accordance containment isolation valve is within with the limits. INSERVICE TESTING PROGRAM SR 3.6.3.6 Perform leakage rate testing for 8 inch In accordance containment purge valves with resilient with the seals. Surveillance Frequency Cont ro 1 Program SR 3.6.3.7 Perform leakage rate testing for 48 inch In accordance containment purge valves with resilient with the seals. Surveillance Frequency Control Program SR 3.6.3.8 Verify each automatic containment isolation In accordance valve that is not locked, sealed or with the otherwise secured in position, actuates to Surveillance the isolation position on an actual or Frequency simulated actuation signal. Control Program BYRON - UNITS 1 & 2 3.6.3 - 6 Amendment 197/197

Containment Spray and Cooling Systems 3.6.6 SURVEILLANCE REQUIREMENTS (continued)

SU RV EI LLANCE FREQUENCY SR 3.6.6.3 Verify each containment cooling train In accordance cooling water flow rate is~ 2660 gpm to with the each cooler. Surveillance Frequency Contra 1 Program SR 3.6.6.4 Verify each containment spray pump's In accordance developed head at the flow test point is with the greater than or equal to the required INSERVICE developed head. TESTING PROGRAM SR 3.6.6.5 Verify each automatic containment s~ray In accordance valve in the flow path that is not ocked, with the sealed, or otherwise secured in position, Surveillance actuates to the correct position on an Frequency actual or simulated actuation signal. Control Program SR 3.6.6.6 Verify each containment s~ray pump starts In accordance automatically on an actua or simulated with the actuation signal. Surveillance Frequency Control Program SR 3.6.6.7 Verify each containment cooling train In accordance starts automatically on an actual or with the simulated actuation signal. Surveillance Frequency Control Program SR 3.6.6.8 Verify each spray nozzle is unobstructed. Following maintenance that could result in nozzle blockage OR Following fluid fl ow through the nozzles (continued)

BYRON - UNITS 1 &2 3.6.6 3 Amendment 197/197

MSSVs 3.7.1 ACTIONS (continued)

B. Required Action and B.l Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND OR B.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> One or more steam generators with ~ 4 MSSVs inoperable.

SURVEILLANCE REQUIREMENTS SU RV EI LLANCE FREQUENCY SR 3. 7 .1.1 -------------------NOTE--------------------

Only required to be performed in MODES 1 and 2.

Verify each required MSSV lift setpoint per In accordance Table 3.7.1-2 in accordance with the with the INSERVICE TESTING PROGRAM. Following INSERVICE testing, lift setting shall be within+/- 1%. TESTING PROGRAM BYRON - UNITS 1 &2 3.7.1-2 Amendment 197/197

MS I Vs 3.7.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 -------------------NOTE--------------------

Only required to be performed in MODES 1 and 2.

Verify closure time of each MSIV is In accordance

5 seconds. with the INSERVICE TESTING PROGRAM SR 3.7.2.2 -------------------NOTE--------------------

Only required to be performed in MODES 1 and 2.

Verify each actuator train actuates the In accordance MSIV to the isolation position on an actual with the or simulated actuation signal. Surveillance Frequency Control Program BYRON - UNITS 1 &2 3.7.2 - 3 Amendment 197/197

AF System 3.7.5 SURVEILLANCE REQUIREMENTS SU RV EI LLANCE FREQUENCY SR 3.7.5.1 Verify each AF manual, power operated, and In accordance automatic valve in each water flow path, with the that is not locked, sealed, or otherwise Surveillance secured in position, is in the correct Frequency position. Control Program SR 3.7.5.2 Verify day tank contains ~ 420 gal of fuel In accordance oi 1. with the Surveillance Frequency Control Program SR 3.7.5.3 Operate the diesel driven AF pump for In accordance

~ 15 minutes. with the Surveillance Frequency Control Program SR 3.7.5.4 Verify the developed head of each AF pump In accordance at the flow test point is greater than or with the equal to the required developed head. INSERVICE TESTING PROGRAM SR 3.7.5.5 Verify each AF automatic valve that is not In accordance locked, sealed, or otherwise secured in with the position, actuates to the correct position Surveillance on an actual or simulated actuation signal. Frequency Control Program SR 3.7.5.6 Verify each AF pump starts automatically on In accordance an actual or simulated actuation signal. with the Surveillance Frequency Control Program (continued)

BYRON - UNITS 1 &2 3.7.5 - 2 Amendment 197/197

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 DELETED BYRON - UNITS 1 &2 5.5 - 6 Amendment 197/197

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT 1 CALVERT CLIFFS NUCLEAR POWER PLANT, LLC EXELON GENERATION COMPANY, LLC DOCKET NO. 50-317 Amendment No. 320 Renewed License No. DPR-53

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated July 26, 2016, as supplemented by letter dated October 6, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations 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; 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 amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-53 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 320, are hereby incorporated into this license. Exelon Generation shall operate the facility in accordance with the Technical Specifications.

Enclosure 5

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

()J 9-David J. Wrona, Branch Chief Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: May 2 6, 2O1 7

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT 2 CALVERT CLIFFS NUCLEAR POWER PLANT, LLC EXELON GENERATION COMPANY. LLC DOCKET NO. 50-318 Amendment No. 298 Renewed License No. DPR-69

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated July 26, 2016, as supplemented by letter dated October 6, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations 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; 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 amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-69 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 298, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications.

Enclosure 6

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION David J. Wrona, Branch Chief Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: May 2 6, 2O1 7

ATTACHMENT TO LICENSE AMENDMENT NOS. 320 AND 298 CALVERT CLIFFS NUCLEAR POWER PLANT, UNITS 1AND2 RENEWED FACILITY OPERATING LICENSE NOS. DPR-53 AND DPR-69 DOCKET NOS. 50-317 AND 50-318 Replace the following pages of the Renewed Facility Operating Licenses and Appendix A, Technical Specifications (TSs), with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove License DPR-53 License DPR-53 Page 3 Page 3 License DPR-69 License DPR-69 Page 3 Page 3 TSs TSs 1.1-3 1.1-3 1.1-4 1.1-4 3.4.10-2 3.4.10-2 3.5.2-2 3.5.2-2 3.6.3-6 3.6.3-6 3.6.6-3 3.6.6-3 3.7.1-2 3.7.1-2 3.7.2-2 3.7.2-2 3.7.3-4 3.7.3-4 3.7.15-1 3.7.15-1 5.5-6 5.5-6 5.5-7 5.5-7 5.5-8 5.5-8 5.5-9 5.5-9 5.5-10 5.5-10 5.5-11 5.5-11 5.5-12 5.5-12 5.5-13 5.5-13 5.5-14 5.5-14 5.5-15 5.5-15 5.5-16 5.5-16 5.5-17 5.5-17 5.5-18 5.5-18 5.5-19 5.5-19 5.5-20

(4) Exelon Generation pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use, in amounts as required, any byproduct, source, and special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Exelon Generation pursuant to the Act and 10 CFR Parts 30 and 70 to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This license is deemed to contain and is subject to the conditions set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act, and the rules, regulations, and orders of the Commission, now or hereafter applicable; and is subject to the additional conditions specified and incorporated below:

(1) Maximum Power Level Exelon Generation is authorized to operate the facility at steady-state reactor core power levels not in excess of 2737 megawatts-thermal in accordance with the conditions specified herein.

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 320 , are hereby incorporated into this license. Exelon Generation shall operate the facility in accordance with the Technical Specifications.

(a) For Surveillance Requirements (SRs) that are new, in Amendment 227 to Facility Operating License No. DPR-53, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 227. For SRs that existed prior to Amendment 227, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment 227.

(3) Additional Conditions The Additional Conditions contained in Appendix C as revised through Amendment No. 318 are hereby incorporated into this license. Exelon Generation shall operate the facility in accordance with the Additional Conditions.

(4) Secondary Water Chemistry Monitoring Program Exelon Generation shall implement a secondary water chemistry monitoring program to inhibit steam generator tube degradation. This program shall include:

Amendment No. 320

(4) Exelon Generation pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use, in amounts as required, any byproduct, source, and special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Exelon Generation pursuant to the Act and 10 CFR Parts 30 and 70 to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This license is deemed to contain and is subject to the conditions set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act, and the rules, regulations, and orders of the Commission, now and hereafter applicable; and is subject to the additional conditions specified and incorporated below:

(1) Maximum Power Level Exelon Generation is authorized to operate the facility at reactor steady-state core power levels not in excess of 2737 megawatts-thermal in accordance with the conditions specified herein.

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 298, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications.

(a) For Surveillance Requirements (SRs) that are new, in Amendment 201 to Facility Operating License No. DPR-69, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 201. For SRs that existed prior to Amendment 201, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment 201.

(3) Less Than Four Pump Operation The licensee shall not operate the reactor at power levels in excess of five (5) percent of rated thermal power with less than four (4) reactor coolant pumps in operation. This condition shall remain in effect until the licensee has submitted safety analyses for less than four pump operation, and approval for such operation has been granted by the Commission by amendment of this license.

(4) Environmental Monitoring Program If harmful effects or evidence of irreversible damage are detected by the biological monitoring program, hydrological monitoring program, and the Amendment No. 298

Definitions 1.1 1.1 Definitions Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion."

£-AVERAGE DISINTEGRATION E shall be the average (weighted in proportion to ENERGY the concentration of each radionuclide in the reactor coolant at the time of sampling} of the sum of the average beta and gamma energies per disintegration (in MeV} for isotopes, other than iodines, with half lives> 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURE The ESF RESPONSE TIME shall be that time interval (ESF} RESPONSE TIME from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e .* the valves travel to their required positions, pump discharge pressures reach their required values. etc.). Times shall include diesel generator starting and sequence loading delays~ where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

INSERVICE TESTING PROGRAM The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

The maximum allowable containment leakage rate, La, shall be 0.16% of containment air weight per day at the calculated peak containment pressure (PJ.

CALVERT CLIFFS - UNIT 1 1.1-3 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 298

Definitions 1.1 1.1 Definitions LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing {except reactor coolant pump (RCP) seal leakoff), that is captured and conducted to collection systems or a sump or collecting tank;

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System {primary to secondary LEAKAGE).

b. Unidentified LEAKAGE All LEAKAGE (except RCP seal leakoff) that is not identified LEAKAGE;

c. Pressure Boundary LEAKAGE LEAKAGE {except primary to secondary LEAKAGE}

through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolts specified in Table 1.1-1 with fuel in the reactor vessel.

CALVERT CLIFFS - UNIT 1 1.1-4 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 298

Pressurizer Safety Valves 3.4.10 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Re qui red Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Reduce all RCS cold 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> leg temperatures to Two pressurizer s 365°F (Unit 1),

safety valves s 301°F (Unit 2).

inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify each pressurizer safety valve is In accordance OPERABLE in accordance with the INSERVICE with the TESTING PROGRAM. The lift settings shall be INSERVICE within limits as specified below: TESTING PROGRAM As Found As Left Valve Lift Setting (psia) Lift Setting (psia)

RC-200 ~ 2475 and s 2575 ~ 2475 and $ 2525 RC-201 ~ 2475 and s 2600 ~ 2500 and s 2550 CALVERT CLIFFS - UNIT 1 3.4.10-2 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 298

ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify the following valves are in the In accordance listed position with power to the valve with the operator removed. Surveillance Frequency Valve Number Position Function Control Program MOV-659 Open Mini-flow Isolation MOV-660 Open Mini-flow Isolation CV-306 Open Low Pressure Safety Injection Flow Control SR 3.5.2.2 -------------------NOTE-------------------

Not required to be met for system vent flow paths opened under administrative control.

Verify each ECCS manual, power-operated, and In accordance automatic valve in the flow path, that is with the not locked, sealed, or otherwise secured in Surveillance position, is in the correct position. Frequency Control Program SR 3.5.2.3 Verify each high pressure safety injection - In accordance and low pressure safety injection pump 1 s with the developed head at the test flow point is INSERVICE greater than or equal to the required TESTING PROGRAM developed head.

SR 3.5.2.4 Deleted CALVERT CLIFFS - UNIT 1 3.5.2-2 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 298

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.3.3 -------------------NOTE-------------------

Val ves and blind flanges in high radiation areas may be verified by use of administrative means.

Verify each containment isolation manual Prior to valve and blind flange that is located entering MODE 4 inside containment and not locked, sealed, from MODE 5 if or otherwise secured and required to be not performed closed during accident conditions is closed, within the except for containment isolation valves that previous are open under administrative controls. 92 days SR 3.6.3.4 Verify the isolation time of each automatic In accordance power-operated containment isolation valve with the is within limits. INSERVICE TESTING PROGRAM SR 3.6.3.5 Verify each automatic containment isolation In accordance valve that is not locked, sealed, or with the otherwise secured in position, actuates to Surveillance the isolation position on an actual or Frequency simulated actuation signal. Control Program CALVERT CLIFFS - UNIT 1 3.6.3-6 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 298

Containment Spray and Cooling Systems 3.6.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1 -------------------NOTE-------------------

Not required to be met for system vent flow paths opened under administrative control.

Verify each containment spray manual, power- In accordance operated, and automatic valve in the flow with the path that is not locked, sealed, or Surveillance otherwise secured in position is in the Frequency correct position. Control Program SR 3.6.6.2 Operate each containment cooling train fan In accordance unit for~ 15 minutes. with the Surveillance Frequency Control Program SR 3.6.6.3 Verify each containment cooling train In accordance cooling water flow rate is~ 2000 gpm to with the each fan cooler. Surveillance Frequency Control Program SR 3.6.6.4 Verify each containment spray pump's In accordance developed head at the flow test point is with the greater than or equal to the required INSERVICE developed head. TESTING PROGRAM SR 3.6.6.5 Verify each automatic containment spray In accordance valve in the flow path that is not locked, with the sealed, or otherwise secured in position, Surveillance actuates to the correct position on an Frequency actual or simulated actuation signal. Control Program CALVERT CLIFFS - UNIT 1 3.6.6-3 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 298

MSSVs 3.7.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> One or more steam generators with less than five MSSVs OPERABLE.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.1.1 -------------------NOTE-------------------

Only required to be performed in MODES 1 and 2.

Verify each required MSSV lift setpoint per In accordance Table 3.7.1-2 in accordance with the with the INSERVICE TESTING PROGRAM. Following INSERVICE testing, lift settings shall be within+/- 1%. TESTING PROGRAM CALVERT CLIFFS - UNIT 1 3.7.1-2 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 298

MS I Vs 3.7.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition C AND not met.

D.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify closure time of each MSIV is within In accordance limits. with the INSERVICE TESTING PROGRAM CALVERT CLIFFS - UNIT 1 3.7.2-2 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 298

AFW System 3.7.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1 Verify each AFW manual, power-operated, and In accordance automatic valve in each water flow path and with the in both steam supply flow paths to the steam Surveillance turbine-driven pumps, that is not locked, Frequency sealed, or otherwise secured in position, is Control Program in the correct position.

SR 3.7.3.2 Cycle each testable, remote-operated valve In accordance that is not in its operating position. with the INSERVICE TESTING PROGRAM SR 3.7.3.3 -------------------NOTE-------------------

Not required to be performed for the turbine-driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 800 psig in the steam generators.

Verify the developed head of each AFW pump In accordance at the flow test point is greater than or with the equal to the required developed head. INS ERV ICE TEST! NG PROGRAM SR 3.7.3.4 -------------------NOTE-------------------

Not required to be performed for the turbine-driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 800 psig in the steam generators.

Verify each AFW automatic valve that is not In accordance locked, sealed, or otherwise secured in with the position, actuates to the correct position Surveillance on an actual or simulated actuation signal. Frequency Control Program CALVERT CLIFFS - UNIT 1 3.7.3-4 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 298

MF I Vs 3.7.15 3.7 PLANT SYSTEMS 3.7.15 Main Feedwater Isolation Valves (MFIVs)

LCO 3.7.15 Two MFIVs shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS

---

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

Separate Condition entry is allowed for each valve.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more MFIVs A.1 Restore MFIV to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable. OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify the closure time of each MFIV is in In accordance accordance with the INSERVICE TESTING with the PROGRAM. IN SERVICE TESTING PROGRAM CALVERT CLIFFS - UNIT 1 3.7.15-1 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 298

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.6 Concrete Containment Tendon Surveillance Program This program provides controls for monitoring any tendon degradation in pre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. The program shall include baseline measurements prior to initial operation. The Tendon Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with Section XI, Subsection IWL of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a, as amended by relief granted in accordance with 10 CFR 50.55a(a)(3).

The provisions of SR 3.0.3 are applicable to the Tendon Surveillance Program inspection frequencies.

5.5.7 Reactor Coolant Pump Flywheel Insoection Program This program shall provide for the inspection of each reactor coolant pump flywheel per the recorrmendations of regulatory position c.4.b of Regulatory Guide 1.14, Revision 1, August 1975.

5.5.8 DELETED 5.5.9 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to CALVERT CLIFFS - UNIT 1 5.5-6 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 298

Programs and Manuals 5.5 5.5 Programs and Manuals the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

b. Performance criteria for SG tube integrity. Steam generator tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down), all anticipated transients included in the design specification, and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady-state full power operation primary to secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 100 gpd per SG.

CALVERT CLIFFS - UNIT 1 5.5-7 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 298

Programs and Manuals 5.5 5.5 Programs and Manuals

3. The operational LEAKAGE performance criterion is specified in LCD 3.4.13, "RCS Operational LEAKAGE."

c. Provisions for SG tube plugging criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial, and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube plugging criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.l, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.

2. After the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, c, and d below. If a degradation CALVERT CLIFFS - UNIT 1 5.5-8 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 298

Programs and Manuals 5.5 5.5 Programs and Manuals assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.

a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months. This constitutes the first inspection period; b) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; c) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d) During the remaining life of the SGs, inspect 100%

of the tubes every 72 effective full power months.

This constitutes the fourth and subsequent inspection periods.

CALVERT CLIFFS - UNIT 1 5.5-9 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 298

Programs and Manuals 5.5 5.5 Programs and Manuals

3. If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspection). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e. Provisions for monitoring operational primary to secondary LEAKAGE.

5.5.10 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit steam generator tube degradation and low pressure turbine disc stress corrosion cracking. The program shall include:

a. Identification of a sampling schedule for the critical variables and control points for these variables;

b. Identification of the procedures used to measure the values of the critical variables;

c. Identification of process sampling points which shall include monitoring the discharge of the condensate pumps for evidence of condenser in leakage;

d. Procedures for the recording and management of data;

e. Procedures defining corrective actions for all off control point chemistry conditions; and

f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of CALVERT CLIFFS - UNIT 1 5.5-10 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 298

Programs and Manuals 5.5 5.5 Programs and Manuals administrative events, which are required to initiate corrective action.

5.5.11 Ventilation Filter Testing Program A program shall be established to implement the following required testing of engineered safety feature (ESF) filter ventilation systems. Tests described in Specifications 5.5.11.a and 5.5.11.b shall be performed once per 18 months for ventilation systems other than the Iodine Removal System {IRS) and 24 months for the IRS; after each complete or partial replacement of the high efficiency particulate air (HEPA) filter bank or charcoal adsorber bank; after any structural maintenance on the HEPA filter or charcoal adsorber housing; and following painting, fire, or chemical release in any ventilation zone communicating with the system.

Tests described in Specification 5.5.11.c shall be performed once per 18 months for ventilation systems other than the IRS and 24 months for the IRS; after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation; after any structural maintenance on the HEPA filter or charcoal adsorber housing; and following painting, fire, or chemical release in any ventilation zone co1m1unicating with the system.

Tests described in Specification 5.5.11.d shall be performed once per 18 months for ventilation systems other than the IRS and 24 months for the IRS.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Ventilation Filter Testing Program test frequencies.

a. Demonstrate for each of the ESF systems that an inplace test of the HEPA filters shows a penetration and system bypass

~ 1.0% (~0.05% for the CREVS only) when tested in accordance with Regulatory Positions C.5.a and C.5.c of Regulatory CALVERT CLIFFS - UNIT 1 5.5-11 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 298

Programs and Manuals 5.5 5.5 Programs and Manuals Guide 1.52, Revision 2, and ANSI N510-1975, at the system flowrate specified as follows+/- 10%:

ESF Ventilation System Flowrate Control Room Emergency Ventilation System 10,000 cfm (CREVS)

Penetration Room Exhaust Ventilation 2,000 cfm System (PREVS)

IRS 20,000 cfm

b. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass s 1.0% when tested in accordance with Regulatory Positions C.5.a and C.5.d of Regulatory Guide 1.52, Revision 2, and ANSI N510-1975, at the system flowrate specified as follows +/- 10%:

ESF Ventilation System Flowrate CREVS 10,000 cfm PREVS 2,000 cfm IRS 20,000 cfm

c. Demonstrate for each of the ESF systems within 31 days after removal that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30°C and greater than or equal to the relative humidity specified as follows:

ESF Ventilation System Penetrations RH CREVS 4.5% 70%

PRE VS 34.5% 95%

IRS 34.5% 95%

CALVERT CLIFFS - UNIT 1 5.5-12 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 298

Programs and Manuals 5.5 5.5 Programs and Manuals

d. For each of the ESF systems, demonstrate the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified below when tested in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1975 at the system flowrate specified as follows +/- 10%:

ESF Ventilation System Delta P Flowrate CREVS 6 inwg 10,000 cfm PREVS 6 inwg 2,000 cfrn IRS 6 inwg 20,000 cfm 5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides control for potentially explosive gas mixtures contained in the Waste Gas Holdup System and the quantity of radioactivity contained in gas storage tanks. The gaseous radioactivity quantities shall be detennined following the methodology in the ODCM.

The program shall include:

a. The limits for concentrations of oxygen in the Waste Gas Holdup System and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion); and

b. A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than or equal to 58,500 curies noble gases (considered as Xe-133).

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance Frequencies.

CALVERT CLIFFS - UNIT 1 5.5-13 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 29 s

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Diesel Fuel Oil Testing Program A Diesel Fuel Oil Testing Program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with ASTM Standards. The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1. An American Petroleum Institute gravity or an absolute specific gravity within limits,

2. A flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and

3. Water and sediment ~ 0.05%.

b. Within 31 days following addition of new fuel oil to the storage tanks, verify that the properties of the new fuel oil, other than those addressed in a., above, are within limits for ASTM 2D fuel oil; and

c. Total particulate concentration of the fuel oil, when determined by gravimetric analysis based on ASTM D2276-1989, is ~ 10 mg/l when tested every 92 days.

d. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Frequencies.

5.5.14 Technical Specifications Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the Technical Specifications shall be made under appropriate administrative controls and reviews.

CALVERT CLIFFS - UNIT 1 5.5-14 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 298

Programs and Manuals 5.5 5.5 Programs and Manuals

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1. A change in the Technical Specifications incorporated in the license; or

2. A change to the UFSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

d. Proposed changes that meet the criteria of Specification 5.5.14b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.7l(e).

5.5.15 Safety Function Determination Program (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into Limiting Condition for Operation (LCO) 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate limitations and remedial or compensatory actions may be identified to be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCD 3.0.6. The SFDP shall contain the following:

a. Provisions for cross-train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected;

b. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists; CALVERT CLIFFS UNIT 1 5.5-15 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 298

Programs and Manuals 5.5 5.5 Programs and Manuals

c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

d. Other appropriate limitations and remedial or compensatory actions.

A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a. A required system redundant to system(s) supported by the inoperable support system is also inoperable; or

b. A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable; or

c. A required system redundant to support system(s) for the supported systems (a) and {b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

5.5.16 Containment Leakage Rate Testing Program A program shall be established to implement the leakage testing of the containment as required by 10 CFR 50.54(0) and 10 CFR Part 50, Appendix J, Option B. This program shall be in accordance with the guidelines contained in Nuclear Energy Institute (NEI) 94-01, "Industry Guideline for Implementing Performance Based Option of 10 CFR Part 50, Appendix J," Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A dated October 2008.

CALVERT CLIFFS - UNIT 1 5.5-16 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 298

Programs and Manuals 5.5 5.5 Programs and Manuals The peak calculated containment internal pressure for the design basis loss-of-coolant accident, Pa* is 49.7 psig. The containment design pressure is 50 psig.

The maximum allowable containment leakage rate, La, shall be 0.16 percent of containment air weight per day at Pa.

Leakage rate acceptance criteria are:

a. Containment leakage rate acceptance criterion is s 1.0 La.

During the first unit startup following testing, in accordance with this program, the leakage rate acceptance criterion are ~ 0.60 La for Types B and C tests and ~ 0.75 La for Type A tests.

b. Air lock testing acceptance criteria are:

1. Overall air lock leakage rate is s 0.05 La when tested at 2 Pa.

2. For each door, leakage rate is ~ 0.0002 La when pressurized to 2 15 psig.

The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

5.5.17 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation System (CREVS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE CALVERT CLIFFS - UNIT 1 5.5-17 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 298

Programs and Manuals 5.5 5.5 Programs and Manuals under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem whole body or its equivalent to any part of the body for the duration of the accident. The program shall include the following elements:

a. The definition of CRE and the CRE boundary.

b. Requirements for maintaining CRE boundary in its design condition including configuration control and preventive maintenance.

c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, 11 Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors,"

Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

d. License controlled programs will be used to verify the integrity of the CRE boundary. Conditions that generate relevant information from those programs will be entered into the corrective action process and shall be trended and used as part of the 36 month assessments of the CRE boundary.

e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered CALVERT CLIFFS - UNIT 1 5.5-18 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 298

Programs and Manuals 5.5 5.5 Programs and Manuals inleakage, and assessing the CRE boundary as required by paragraphs c and d respectively.

5.5.18 Not Used 5.5.19 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NE! 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk Informed Method for Control of Surveillance Frequencies," Revision 1.

c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

CALVERT CLIFFS - UNIT 1 5.5-19 Amendment No. 320 CALVERT CLIFFS - UNIT 2 Amendment No. 298

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY. LLC DOCKET NO. 50-461 CLINTON POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 212 License No. NPF-62

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated July 26, 2016, as supplemented by letter dated October 6, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations 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; 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 amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-62 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 212, are hereby incorporated into this license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Enclosure 7

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days ()f the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Q~na~;anc~'-ie_f Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Facility Operating License Date of Issuance: May 2 6 , 2O1 7

ATTACHMENT TO LICENSE AMENDMENT NO. 212 CLINTON POWER STATION, UNIT NO. 1 FACILITY OPERATING LICENSE NO. NPF-62 DOCKET NO. 50-461 Replace the following pages of the Facility Operating License and Appendix A, Technical Specifications (TSs), with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove License NPF-62 License NPF-62 Page 3 Page 3 TSs TSs 1.0-3 1.0-3 3.1-22 3.1-22 3.4-11 3.4-11 3.4-16 3.4-16 3.5-4 3.5-4 3.5-9 3.5-9 3.6-18 3.6-18 3.6-25 3.6-25 3.6-33 3.6-33 3.6-65 3.6-65 5.0-11 5.0-11

(4) Exelon Generation Company, pursuant to the Act and to 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5) Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (6) Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

Mechanical disassembly of the GE14i isotope test assemblies containing Cobalt-60 is not considered separation; and (7) Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30, to intentionally produce, possess, receive, transfer, and use Cobalt-60.

C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level Exelon Generation Company is authorized to operate the facility at reactor core power levels not in excess of 3473 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 212 , are hereby incorporated into this license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Amendment No. 212

Definitions 1.1 1.1 Definitions (continued)

EMERGENCY CORE COOLING The ECCS RESPONSE TIME shall be that time interval SYSTEM (ECCS) RESPONSE from when the monitored parameter exceeds its ECCS TIME initiation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

END OF CYCLE The EOC-RPT SYSTEM RESPONSE TIME shall be that RECIRCULATION PUMP TRIP time interval from initial movement of the (EOC-RPT) SYSTEM RESPONSE associated turbine stop valve or turbine TIME control valve to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of lC CFR 50.55a(f).

ISOLATION SYSTEM The ISOLATION SYSTEM RESPONSE TIME shall be that RESPONSE TIME time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves travel to their required positions. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

(continued)

CLINTON .a Amer:dment 'lo. 212

SLC System

3. 1. 7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.7.7 Verify each pump develops a flow rate In accordance

~ 41.2 gprn at a discharge pressure with the

~ 1220 psig.

INSERVICE TESTING PROGRAM SR 3. 1. 7. 8 Verify flow through one SLC subsystem from In accordance pump into reactor pressure vessel. with the Surveillance Frequency Control Program SR 3.1.7.9 Verify all piping between storage tank and In accordance pump suction is unblocked. with the Surveillance Frequency Control Program AND Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after pump suction piping temperature is restored to

~ 70°F CLINTON 3 .. -22 }l.mendrnent No. 212

S/RVs

3. 4. 4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify the safety function lift setpoints In accordance of the required S/RVs are as follows: with the INSERVICE Number of Setpoint TESTING PROGRAM S/RVs (psig) 7 1165 +/- 34.9 5 1180 +/- 35.4 4 1190 +/- 35. 7 Following testing, lift settings shall be within +/- 1%.

SR 3.4.4.2 -------------------NOTE-------------------

Val ve actuation may be excluded.

Verify each required relief function S/RV In accordance actuates on an actual or simulated with the automatic initiation signal. Surveillance Frequency Control Program SR 3.4.4.3 -------------------NOTE-------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify each required S/RV actuator strokes In accordance when manually actuated. with the Surveillance Frequency Control Program CLINTON 3.4-11 Amendment No. 212

RCS PIV Leakage

3. 4. 6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 -------------------NOTE-------------------

Not required to be performed in MODE 3.

Verify equivalent leakage of each RCS PIV In accordance is~ 0.5 gpm per nominal inch of valve size with the up to a maximum of 5 gpm, at an RCS INSERVICE pressure ~ 1000 psig and ~ 1025 psig. TEST ING PROGRAM CLINTON 3.4-16 l\fl\endment No. 212

ECCS-Operating

3. 5 .1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3. 5 .1.1 Verify, for each ECCS injection/spray In accordance subsystem, locations susceptible to gas with the accumulation are sufficiently filled with Surveillance water. Frequency Control Program SR 3. 5 .1.2 -------------------NOTES-------------------

1. Low pressure coolant injection (LPCI) subsystems may be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the residual heat removal cut in permissive pressure in MODE 3, if capable of being manually realigned and not otherwise inoperable.

2. Not required to be met for system vent flow paths opened under administrative control.

Verify each ECCS injection/spray subsystem In accordance manual, power operated, and automatic valve with the in the flow path, that is not locked, Surveillance sealed, or otherwise secured in position, Frequency is in the correct position. Control Program SR 3. 5. 1. 3 Verify ADS accwnulator supply pressure is In accordance

140 psig. with the Surveillance Frequency Control Program SR 3. 5 .1. 4 Verify each ECCS pump develops the In accordance specified flow rate with the specified pump with the differential pressure. IN SERVICE TESTING PROGRAM PUMP DIFFERENTIAL SYSTEM FLOW RATE PRESSURE LPCS  ;::: 5010 gpm ~ 290 psid LPCI  ;::: 5050 gpm ~ 113 psid HPCS  ;:: 5010 gprn ~ 363 psid (continued)

CLINTON 3.5-4 Arnendment No. 212

ECCS-Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE E'REQUENCY SR 3.5.2.5 Verify each required ECCS pump develops the In accordance specified flow rate with the specified pwnp with the differential pressure. INSERVICE TESTING PROGRAM PUMP DIFFERENTIAL SYSTEM FLOW RATE PRESSURE LPCS  ;::.: 5010 gpm ~ 290 psid LPCI  ;::.: 5050 gpm ~ 113 psid HPCS  ;::.: 5010 gpm  ;::.: 363 psid SR 3.5.2.6 -------------------NOTE--------------------

Ves sel injection/spray may be excluded.

Verify each required ECCS injection/spray In accordance subsystem actuates on an actual or with the simulated automatic initiation signal. Surveillance Frequency Control Program CLINTON 3.5-9 Amendment No. 212

PC I Vs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3. 6. 1. 3. 4 Verify the isolation time of each power In accordance operated and each automatic PCIV, except with the MSIVs, is within limits. IN SERVICE TESTING PROGRAM SR 3.6.1.3.5 ------------------NOTE------------------

Only required to be met in MODES 1, 2, and 3.

Perform leakage rate testing for each Once within 92 primary containment purge valve with days after resilient seals. opening the valve AND In accordance with the Primary Containment Leakage Rate Testing Program SR 3. 6. 1. 3. 6 Verify the isolation time of each MSIV In accordance is ~ 3 seconds and~ 5 seconds. with the INSERVICE TESTING PROGRAM SR 3. 6. 1. 3. 7 Verify each automatic PCIV actuates to In accordance the isolation position on an actual or with the simulated isolation signal. Surveillance Frequency Control program (continued)

CLINTON 3.6-18 Amendment No. 212

RHR Containment Spray System 3.6.1.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.7.1 ------------------NOTES------------------

1. RHR containment spray subsystems may be considered OPERABLE during alignment and operation for decay heat removal when below the RHR cut in permissive pressure in MODE 3 if capable of being manually realigned and not otherwise inoperable.

2. Not required to be met for system vent flow paths opened under administrative control.

Verify each RHR containment spray In accordance subsystem manual, power operated, and with the automatic valve in the flow path that is Surveillance not locked, sealed, or otherwise secured Frequency in position is in the correct position. Control Program SR 3.6.1.7.2 Verify each RHR pump develops a flow rate In accordance of ~ 3800 gpm on recirculation flow with the through the associated heat exchanger to INSERVICE the suppression pool. TESTING PROGRAM SR 3.6.1.7.3 Verify each RHR containment spray In accordance subsystem automatic valve in the flow with the path actuates to its correct position on Surveillance an actual or simulated automatic Frequency initiation signal. Control Program SR 3.6.1.7.4 Verify each spray nozzle is unobstructed. Following activities that could result in nozzle blockage SR 3.6.1.7.5 Verify RHR containment spray subsystem In accordance locations susceptible to gas accumulation with the are sufficiently filled with water. Surveillance Frequency Control Program CLINTON 3.6-25 Amendment No. 212

RHR Suppression Pool Cooling

3. 6.2. 3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.3.l Verify each RHR suppression pool cooling In accordance subsystem manual, power operated, and with the automatic valve in the flow path that is Surveillance not locked, sealed, or otherwise secured Frequency in position is in the correct position or Control Program can be aligned to the correct position.

SR 3.6.2.3.2 Verify each RHR pump develops a flow rate In accordance

~ 4550 gprn through the associated heat with the exchanger to the suppression pool. IN SERVICE TESTING PROGRAM SR 3.6.2.3.3 Verify RHR suppression pool cooling In accordance subsystem locations susceptible to gas with the accumulation are sufficiently filled with Surveillance water. Frequency Control Program CLINTON 3.6-33 Amen~~ent No. 212

Drywell Isolation Valves

3. 6 .5. 3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.5.3.3 ------------------NOTES------------------

1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2. Not required to be met for drywell isolation valves that are open under administrative controls.

Verify each required drywell isolation Prior to manual valve and blind flange that is entering MODE 2 required to be closed during accident or 3 from conditions is closed. MODE 4, if not performed in the previous 92 days SR 3.6.5.3.4 Verify the isolation time of each In accordance required power operated and each required with the automatic drywell isolation valve is IN SERVICE within limits. TESTING PROGRAM SR 3.6.5.3.5 Verify each required automatic drywell In accordance isolation valve actuates to the isolation with the position on an actual or simulated Surveillance isolation signal. Frequency Control Program CLINTON 3.6-65 AmencLrnent No. 212

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.5 Component Cyclic or Transient Limit This program provides controls to track the cyclic and transient occurrences identified on USAR Table 3.9-l(b) to ensure that the reactor vessel is maintained within the design limits.

5.5.6 DELETED (continued)

CLINTON 5.0-11 Amendment No. 212

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY. LLC DOCKET NO. 50-237 DRESDEN NUCLEAR POWER STATION. UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 254 Renewed License No. DPR-19

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated July 26, 2016, as supplemented by letter dated October 6, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations 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; 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 amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-19 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 254, are hereby incorporated into this renewed operating license.

The licensee shall operate the facility in accordance with the Technical Specifications.

Enclosure 8

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Q~na,~;anc~:;-

Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: May 26, 2017

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY. LLC DOCKET NO. 50-249 DRESDEN NUCLEAR POWER STATION. UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 247 Renewed License No. DPR-25

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated July 26, 2016, as supplemented by letter dated October 6, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations 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; 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 amended by 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-25 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 24 7, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

Enclosure 9

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

()J 9*

David J. Wrona, Branch Chief rV ' -

Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: May 2 6, 2O1 7

ATTACHMENT TO LICENSE AMENDMENT NOS. 254 AND 247 DRESDEN NUCLEAR POWER STATION. UNITS 2 AND 3 RENEWED FACILITY OPERATING LICENSE NOS. DPR-19 AND DPR-25 DOCKET NOS. 50-237 AND 50-249 Replace the following pages of the Renewed Facility Operating Licenses and Appendix A, Technical Specifications (TSs), with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove License DPR-19 License DPR-19 Page 3 Page 3 License DPR-25 License DPR-25 Page4 Page 4 TSs TSs 1.1-3 1.1-3 3.1.7-3 3.1.7-3 3.4.3-2 3.4.3-2 3.5.1-4 3.5.1-4 3.5.1-5 3.5.1-5 3.5.2-4 3.5.2-4 3.6.1.3-8 3.6.1.3-8 3.6.2.3-2 3.6.2.3-2 5.5-4 5.5-4 5.5-5 5.5-5

(2) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear materials as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Updated Final Safety Analysis Report, as supplemented and amended; (3) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct special nuclear materials as may be produced by the operation of the facility.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2957 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 254 , are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Operation in the coastdown mode is permitted to 40% power.

Renewed License No. DPR-19 Amendment No. 254 I

f. Surveillance Requirement 4.9.A.10 - Diesel Storage Tank Cleaning (Unit 3 and Unit 2/3 only)

Each of the above Surveillance Requirements shall be successfully demonstrated prior to entering into MODE 2 on the first plant startup following the fourteenth refueling outage (D3R14).

3. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

A. Maximum Power Level The licensee is authorized to operate the facility at steady state power levels not in excess of 2957 megawatts (thermal), except that the licensee shall not operate the facility at power levels in excess of five (5) megawatts (thermal), until satisfactory completion of modifications and final testing of the station output transformer, the auto-depressurization interlock, and the feedwater system, as described in the licensee's telegrams; dated February 26, 1971, have been verified in writing by the Commission.

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 247. are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

C. Reports The licensee shall make certain reports in accordance with the requirements of the Technical Specifications.

D. Records The licensee shall keep facility operating records in accordance with the requirements of the Technical Specifications.

E. Restrictions Operation in the coastdown mode is permitted to 40% power.

Renewed License No. DPR-25 Amendment No. 247

Definitions 1.1 1.1 Definitions DOSE EQUIVALENT J-131 Guidance Report 11, "Limiting Values of (continued) Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989.

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE into the drywel 1 , such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or

2. LEAKAGE into the drywel l atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;

b. Unidentified LEAKAGE A11 LEAKAGE into the drywel l that is not identified LEAKAGE;

c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE; and

d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

(continued)

Dresden 2 and 3 1.1 *3 Amendment No. 254/247

SLC System 3 .1. 7 SURVEILLANCE REQUIREMENTS SU RV EI LLANC E FREQUENCY SR 3.1.7.5 Verify the concentration of sodium In accordance pentaborate in solution is within the with the limits of Figure 3.1.7-1. Surveillance Frequency Control Program Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after water or sodium pentaborate is added to solution Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after solution temperature is restored within the 1 i mi ts of Figure 3.1.7-2 SR 3.1.7.6 Verify each SLC subsystem manual valve in In accordance the flow path that is not locked, sealed, with the or otherwise secured in position is in the Surveillance correct position, or can be aligned to the Frequency correct position. Control Program SR 3.1.7.7 Verify each pump develops a flow rate In accordance

~ 40 gpm at a discharge pressure with the

~ 1275 psig. INS ERV ICE TESTING PROGRAM SR 3.1.7.8 Verify flow through one SLC subsystem from In accordance pump into reactor pressure vessel. with the Surveillance Frequency Control Program (continued)

Dresden 2 and 3 3.1.7-3 Amendment No. 254/247

Safety and Relief Valves 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift setpoints In accordance of the safety valves are as follows: with the INSERVICE Number of Setpoint TESTING PROGRAM Safety Valves (psig) 1 1135 +/- 34 .1 2 1240 +/- 37.2 2 1250 +/- 37.5 4 1260 +/- 37.8 Following testing, lift settings shall be within +/- 1%.

SR 3.4.3.2 Verify each relief valve actuator strokes In accordance when manually actuated. with the Surveillance Frequency Control Program SR 3.4.3.3 - - -- - - - - - - - - - - - - - - -NOTE- - -- - - - - - - - - - - - - - -- -

Valve actuation may be excluded.

Verify each relief valve actuates on an In accordance actual or simulated automatic initiation with the signal. Surveillance Frequency Control Program Dresden 2 and 3 3.4.3-2 Amendment No. 254/247

ECCS-Operating

3. 5 .1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.l Verify, for each ECCS injection/spray In accordance subsystem, locations susceptible to gas with the accumulation are sufficiently filled with Surveillance water. Frequency Control Program SR 3.5.1.2 - - - - - - - - - - - - - - - - - - - -NOTE- - - - - - - - - - - - - - - - -

Not required to be met for system vent flow paths opened under administrative control.

Verify each ECCS injection/spray subsystem In accordance manual, power operated, and automatic valve with the in the flow path, that is not locked, Surveillance sealed, or otherwise secured in position, Frequency is in the correct position. Control Program SR 3.5.1.3 Verify correct breaker alignment to the In accordance LPCI swing bus. with the Surveillance Frequency Control Program SR 3.5.1.4 Verify each recirculation pump discharge In accordance valve cycles through one complete cycle of with the full travel or is de-energized in the INSERVICE closed position. TESTING PROGRAM SR 3.5.1.5 Verify the following ECCS pumps develop the In accordance specified flow rate against a test line with the pressure corresponding to the specified INS ERV ICE reactor pressure. TESTiNG PROGRAM TEST LINE PRESSURE NO. CORRESPONDING OF TO A REACTOR SYSTEM FLOW RATE £Jlli£..S. PRESSURE OF Core Spray ~ 4500 gpm 1 ~ 90 psig LPCI ~ 9000 gpm 2 ~ 20 psig (continued)

Dresden 2 and 3 3. 5 .1-4 Amendment No. 254/247

ECCS-Operating 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.6 - - - - - - - - - - - - - - - - - - - NOTE- - - - - - - - - - - - - - - - - - - -

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure ~ 1005 and In accordance

~ 920 psig, the HPCI pump can develop a with the flow rate~ 5000 gpm against a system head INSERVICE corresponding to reactor pressure. TESTING PROGRAM SR 3.5.1.7 - - - - - - - - - - - - - - - - - - -NOTE- - - - - - - - - - - - - - - - - - - -

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure~ 180 psig, In accordance the HPCI pump can develop a flow rate with the

~ 5000 gpm against a system head Surveillance corresponding to reactor pressure. Frequency Control Program SR 3.5.1.8 - - - - - - - - - - - - - - - - - - - NOTE- - - - - - - - - - - - - - - - - - - -

Vessel injection/spray may be excluded.

Verify each ECCS injection/spray subsystem In accordance actuates on an actual or simulated with the automatic initiation signal. Surveillance Frequency Control Program (continued)

Dresden 2 and 3 3.5.1-5 Amendment No.254/247

ECCS- Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.4 Verify each required ECCS pump develops the In accordance specified flow rate against a test line with the pressure corresponding to the specified INSERVICE reactor pressure. TESTING PROGRAM TEST LINE PRESSURE NO. CORRESPONDING OF TO A REACTOR SYSTEM FLOW RATE .PJll1E.5_ PRESSURE OF cs  ;;;: 4500 gpm 1  ;;;: 90 psig LPCI ~ 4500 gpm 1  ;;;: 20 psig SR 3.5.2.5 - - - - - - - - - - - - - - - - - - -NOTE- - - - - - - - - - - - - - - - - - - -

Vessel injection/spray may be excluded.

Verify each required ECCS injection/spray In accordance subsystem actuates on an actual or with the simulated automatic initiation signal. Surveillance Frequency Control Program Dresden 2 and 3 3.5.2-4 Amendment No. 254/247

PCI Vs 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.4 Verify continuity of the traversing In accordance incore probe (TIP) shear isolation valve with the explosive charge. Surveillance Frequency Control Program SR 3. 6. 1. 3. 5 Verify the isolation time of each power In accordance operated, automatic PCIV, except for with the MSIVs, is within limits. iNSERVICE TESTING PROGRAM SR 3.6.1.3.6 Verify the isolation time of each MSIV is In accordance

~ 3 seconds and ~ 5 seconds. with the INSERVICE TESTING PROGRAM SR 3.6.1.3.7 Verify each automatic PCIV actuates to In accordance the isolation position on an actual or with the simulated isolation signal. Surveillance Frequency Control Program SR 3.6.1.3.8 Verify a representative sample of reactor In accordance instrumentation line EFCVs actuate to the with the isolation position on an actual or Surveillance simulated instrument line break signal. Frequency Control Program SR 3.6.1.3.9 Remove and test the explosive squib from In accordance each shear isolation valve of the TIP with the System. Surveillance Frequency Control Program (continued)

Dresden 2 and 3 3.6.1.3-8 Amendment No. 254/247

Suppression Pool Cooling 3.6.2.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.3.1 Verify each suppression pool cooling In accordance subsystem manual and power operated valve with the in the flow path that is not locked, Surveillance sealed, or otherwise secured in position, Frequency is in the correct position or can be Control Program aligned to the correct position.

SR 3.6.2.3.2 verify each required LPCI pump develops a In accordance flow rate~ 5000 gpm through the with the associated heat exchanger while operating INSERVICE in the suppression pool cooling mode. TESTING PROGRAM SR 3.6.2.3.3 Verify suppression pool cooling subsystem In accordance locations susceptible to gas accumulation with the are sufficiently filled with water. Surveillance Frequency Control Program Dresden 2 and 3 3.6.2.3-2 Amendment No. 254/247

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Raqioactjve Effluent Controls Program (continued)

1. For noble gases: a dose rate~ 500 mrems/yr to the whole body and a dose rate $ 3000 mrems/yr to the skin, and

2. For iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days: a dose rate $ 1500 mrems/yr to any organ;

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I;

i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives> 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and

j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluents Control Program Surveillance Frequencies.

5.5.5 Component Cvclic or Transient Limit This program provides controls to track the UFSAR Section 3.9, cyclic and transient occurrences to ensure that components are maintained within the design limits.

5. 5. 6 DELETED (continued)

Dresden 2 and 3 5.5-4 Amendment No. 254/247

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VETP)

The VFTP shall establish the required testing of Engineered Safety Feature {ESF) filter ventilation systems. Tests described in Specification 5.5.7.a and 5.5.7.b shall be performed once per 24 months; after each complete or partial replacement of the HEPA filter bank or charcoal adsorber bank; after any structural maintenance on the HEPA filter bank or charcoal adsorber bank housing; and, following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation that could adversely affect the filter bank or charcoal adsorber capability.

(continued)

Dresden 2 and 3 5.5-5 Amendment No. 254/247

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. 50-373 LASALLE COUNTY STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 223 Renewed License No. NPF-11

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated July 26, 2016, as supplemented by letter dated October 6, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations 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; 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 amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-11 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 223, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Enclosure 10

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION C)J 2 ~;__

David J. Wrona, Branch Chief Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: May 2 6, 2O1 7

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY. LLC DOCKET NO. 50-37 4 LASALLE COUNTY STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 209 Renewed License No. NPF-18

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated July 26, 2016, as supplemented by letter dated October 6, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations 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; 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 amended by changes to the Technical Specifications as indicated in the enclosure to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-18 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 209, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Enclosure 11

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION C)J 9 Y____

David J. Wrona, Branch Chief Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: May 2 6 , 2O1 7

ATTACHMENT TO LICENSE AMENDMENT NOS. 223 AND 209 LASALLE COUNTY STATION, UNITS 1 AND 2 FACILITY OPERATING LICENSE NOS. NPF-11 AND NPF-18 DOCKET NOS. 50-373 AND 50-374 Replace the following pages of the Renewed Facility Operating Licenses and Appendix A, Technical Specifications (TSs), with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove License NPF-11 License NPF-11 Page 3 Page 3 License NPF-18 License NPF-18 Page 3 Page 3 TSs TSs 1.1-4 1.1-4 3.1.7-3 3.1.7-3 3.4.4-2 3.4.4-2 3.4.6-3 3.4.6-3 3.5.1-5 3.5.1-5 3.5.2-3 3.5.2-3 3.6.1.3-7 3.6.1.3-7 3.6.1.3-8 3.6.1.3-8 3.6.2.3-2 3.6.2.3-2 3.6.2.4-2 3.6.2.4-2 5.5-5 5.5-5 5.5-6 5.5-6

Renewed License No. NPF-11 (3) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; Am. 146 (4) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 01/12/01 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and Am. 202 (5) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 07/21/11 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of LaSalle County Station, Units 1 and 2, and such Class Band Class Clow-level radioactive waste as may be produced by the operation of Braidwood Station, Units 1 and 2, Byron Station, Units 1 and 2, and Clinton Power Station, Unit 1.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

Am. 198 (1) Maximum Power Level 09/16/10 The licensee is authorized to operate the facility at reactor core power levels not in excess of full power (3546 megawatts thermal).

Am. 223 (2) Technical Specifications and Environmental Protection Plan 05/26/17 The Technical Specifications contained in Appendix A, as revised through Amendment No. 223, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license.

The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Am. 194 (3) DELETED 08/28/09 Am. 194 (4) DELETED 08/28/09 Am. 194 (5) DELETED 08/28/09 Amendment No. 223

Renewed License No. NPF-18 (2) Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and Am. 189 (5) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 07/21/11 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of LaSalle County Station, Units 1 and 2, and such Class B and Class C low-level radioactive waste as may be produced by the operation of Braidwood Station, Units 1 and 2, Byron Station, Units 1 and 2, and Clinton Power Station, Unit 1.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

Am. 185 (1) Maximum Power Level 09/16/10 The licensee is authorized to operate the facility at reactor core power levels not in excess of full power (3546 megawatts thermal). Items in Attachment 1 shall be completed as specified. Attachment 1 is hereby incorporated into this license.

Am. 209 (2) Technical Specifications and Environmental Protection Plan 05/26/17 The Technical Specifications contained in Appendix A, as revised through Amendment No. 209, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license.

The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Amendment No. 209

Definitions 1.1 1.1 Definitions (continued)

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).

ISOLATION SYSTEM The ISOLATION SYSTEM RESPONSE TIME shall be that RESPONSE TIME time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves travel to their required positions. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and method for verification have been previously reviewed and approved by the NRC.

LEAKAGE LEAKAGE shal 1 be:

a. Identified LEAKAGE

1. LEAKAGE into the drywel l such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or

2. LEAKAGE into the drywel 1 atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;

b. Unidentified LEAKAGE A11 LEAKAGE into the drywel 1 that is not identified LEAKAGE;

c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE; and

d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System CRCS) component body, pipe wall, or vessel wall.

(continued)

LaSalle 1 and 2 1. 1-4 Amendment No. 223/209

SLC System 3 .1. 7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.5 Verify the concentration of sodium In accordance pentaborate in solution is within the with the limits of Figure 3.1.7-1. Surveillance Frequency Control Program Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after water or sodium pentaborate is added to solution Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after solution temperature is restored within the 1 i mits of Figure 3.1.7-2 SR 3.1.7.6 Verify each SLC subsystem manual, power In accordance operated, and automatic valve in the flow with the path that is not locked, sealed, or Surveillance otherwise secured in position is in the Frequency correct position, or can be aligned to the Control Program correct position.

SR 3.1.7.7 Verify each pump develops a flow rate In accordance

~ 41.2 gpm at a discharge pressure with the

~ 1220 psig. INS ERV ICE TESTING PROGRAM (continued)

LaSalle 1 and 2 3.1.7-3 Amendment No. 223/209

S/RVs 3.4.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 - - - - -- - - - - - - - - - - - - - -NOTE- - - - - - - - - - - - - - - - - - -

Less than or equal to two required S/RVs may be changed to a lower setpoint group.

Verify the safety function lift setpoints In accordance of the required S/RVs are as follows: with the INSERVICE Number of Setpoint TESTING PROGRAM

$/RVs (psigl 2 1205 +/- 36.1 3 1195 +/- 35.8 2 1185 +/- 35. 5 4 1175+/-35.2 2 1150 +/- 34.5 Following testing, lift settings shall be within +/- 1%.

LaSalle l and 2 3.4.4-2 Amendment No. 223/209

RCS PIV Leakage 3.4.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 - - - - - - - - - - - - - - - - - - -NOTE - - - - - - - - - - - - - - - - - - - -

Only required to be performed in MODES 1 and 2.

Verify equivalent leakage of each RCS PIV In accordance is~ 0.5 gpm per nominal inch of valve size with the up to a maximum of 5 gpm, at an RCS INSERVICE pressure ~ 950 psig and ~ 1050 psig. TESTING PROGRAM LaSalle 1 and 2 3.4.6-3 Amendment No. 223/209

ECCS-Operating

3. 5 .1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.5 Verify each ECCS pump develops the In accordance specified flow rate against the specified with the test line pressure. INSERVICE TESTING PROGRAM TEST LINE SY SIEM FLOW RATE PRESSURE LPCS  ?: 6350 gpm ~ 290 psig LPCI 2: 7200 gpm  ?: 130 psig HPCS (Unit ll ~ 6250 gpm 2: 370 psig HPCS (Unit 2) 2: 6200 gpm  ?: 330 psig SR 3.5.1.6 - - - - - - - - - - - - - - - - - - -NOTE - - - - - - - - - - - - - - - - - - - -

Vessel injection/spray may be excluded.

Verify each ECCS injection/spray subsystem In accordance actuates on an actual or simulated with the automatic initiation signal. Surveillance Frequency Control Program SR 3.5.1.7 - - - - - - - - - - - - - - - - - - -NOTE- - - - - - - - - - - - - - - - - - - -

Valve actuation may be excluded.

Verify the ADS actuates on an actual or In accordance simulated automatic initiation signal. with the Surveillance Frequency Control Program SR 3.5.1.8 - - - - - - - - - - - - - - - - - - - NOTE- - - - - - - - - - - - - - - - - - - -

Valve actuation may be excluded.

Verify each required ADS valve actuator In accordance strokes when manually actuated. with the Surveillance Frequency Control Program LaSalle 1 and 2 3.5.1-5 Amendment No. 223/209

ECCS-Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify, for each required low pressure ECCS In accordance injection/spray subsystem, the suppression with the pool water level is?. -12 ft 7 in. Surveillance Frequency Control Program SR 3.5.2.2 Verify, for the required High Pressure Core In accordance Spray (HPCS) System, the suppression pool \vi th the water level is?. -12 ft 7 in. Surveillance Frequency Control Program SR 3.5.2.3 Verify, for each required ECCS injection/ In accordance spray subsystem, locations susceptible to gas with the accumulation are sufficiently filled with Surveillance water. Frequency Control Program SR 3.5.2.4 - - - - - - - - - - - - - - - - - - - NOTE- - - - - - - - - - - - - - - - - - - -

Not required to be met for system vent flow paths opened under administrative control.

Verify each required ECCS injection/spray In accordance subsystem manual, power operated, and with the automatic valve in the flow path, that is Surveillance not locked, sealed, or otherwise secured in Frequency position, is in the correct position. Control Program SR 3.5.2.5 Verify each required ECCS pump develops the In accordance specified flow rate against the specified with the test line pressure. INSERVICE TESTING PROGRAM TEST LINE SYSTEM FLOW RATE PRESSURE LPCS  ?. 6350 gpm  ?. 290 psig LPCI  ?. 7200 gpm  ?. 130 psig HPCS (Unit 1)  ?. 5250 gpm  ?. 370 psig HPCS CUnit 2)  ?. 6200 gpm ~ 330 psig (continued)

LaSalle 1 and 2 3.5.2-3 Amendment No. 223/209

PC I Vs 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.3 - - - - - - - - - - - - - - - - - -NOTES- - - - - - - - - - - - - - - - - -

1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2. Not required to be met for PCIVs that are open under administrative controls.

Verify each primary containment isolation Prior to manual valve and blind flange that is entering MODE 2 located inside primary containment and or 3 from not locked, sealed, or otherwise secured MODE 4 if and is required to be closed during primary accident conditions is closed. containment was de-inerted while in MODE 4, i f not performed within the previous 92 days SR 3.6.1.3.4 Verify continuity of the traversing In accordance incore probe (TIP) shear isolation valve with the explosive charge. Surveillance Frequency Control Program SR 3.6.1.3.5 Verify the isolation time of each power In accordance operated, automatic PCIV, except MSIVs, with the is within limits. INSERVICE TESTING PROGRAM (continued)

LaSalle 1 and 2 3.6.1.3-7 Amendment No. 223/209

PC I Vs 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.6 Verify the isolation time of each MSIV is In accordance

~ 3 seconds and ~ 5 seconds. with the INSERVICE TES TI NG PROGRAM SR 3.6.1.3.7 Verify each automatic PCIV actuates to In accordance the isolation position on an actual or with the simulated isolation signal. Surveillance Frequency Control Program SR 3.6.1.3.B Verify each reactor instrumentation line In accordance EFCV actuates to the isolation position with the on an actual or simulated instrument line Surveillance break signal. Frequency Control Program SR 3.6.1.3.9 Remove and test the explosive squib from In accordance each shear isolation valve of the TIP with the System. Surveillance Frequency Control Program SR 3.6.1.3.10 Verify leakage rate through any one main In accordance steam line is$ 200 scfh and through all with the four main steam lines is~ 400 scfh when Primary tested at ~ 25.0 psig. Containment Leakage Rate Testing Program SR 3.6.1.3.11 Verify combined leakage rate through In accordance hydrostatically tested lines that with the penetrate the primary containment is Primary within limits. Containment Leakage Rate Testing Program LaSalle 1 and 2 3.6.1.3-8 Amendment No. 223/209

RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCE REQUIREMENTS SU RV EI LLANCE FREQUENCY SR 3.6.2.3.1 Verify each RHR suppression pool cooling In accordance subsystem manual and power operated valve with the in the flow path that is not locked, Surveillance sealed, or otherwise secured in position, Frequency is in the correct position or can be Control Program aligned to the correct position.

SR 3.6.2.3.2 Verify each required RHR pump develops a In accordance flow rate~ 7200 gpm through the with the associated heat exchanger while operating INSERVICE in the suppression pool cooling mode. TESTING PROGRAM SR 3.6.2.3.3 Verify RHR suppression pool cooling In accordance subsystem locations susceptible to gas with the accumulation are sufficiently filled with Surveillance water. Frequency Control Program LaSalle 1 and 2 3.6.2.3-2 Amendment No. 223/209

RHR Suppression Pool Spray 3.6.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.4.1 Verify each RHR suppression pool spray In accordance subsystem manual and power operated valve with the in the flow path that is not locked, Surveillance sealed, or otherwise secured in position, Frequency is in the correct position or can be Control Program aligned to the correct position.

SR 3.6.2.4.2 Verify each required RHR pump develops a In accordance flow rate~ 450 gpm through the spray with the sparger while operating in the INSERVICE suppression pool spray mode. TESTING PROGRAM SR 3.6.2.4.3 Verify RHR suppression pool spray In accordance subsystem locations susceptible to gas with the accumulation are sufficiently filled with Surveillance water. Frequency Control Program LaSalle 1 and 2 3.6.2.4-2 Amendment No. 223/209

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.6 Inservice Inspection Program for Post Tensioning Tendons This program provides controls for monitoring any tendon degradation in pre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. The program shall include baseline measurements prior to initial operations. The Tendon Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with Section XI, Subsection IWL of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a as amended by relief granted in accordance with 10 CFR 50.55a(a)(3).

The provisions of SR 3.0.3 are applicable to the Tendon Surveillance Program inspection frequencies.

5.5.7 DELETED (continued)

LaSalle 1 and 2 5.5-5 Amendment No. 223/209

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Ventilgtion Filter Testing Program CVFTP)

The VFTP shall establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems. Tests described in Specification 5.5.8.a and 5.5.8.b shall be performed once per 24 months; after each complete or partial replacement of the HEPA filter bank or charcoal adsorber bank; after any structural maintenance on the HEPA filter bank or charcoal adsorber bank housing; and, following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation that could adversely affect the filter bank or charcoal adsorber capability.

Tests described in Specification 5.5.8.c shall be performed once per 24 months; after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation; after any structural maintenance on the charcoal adsorber bank housing; and, following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation that could adversely affect the charcoal adsorber capability.

Tests described in Specification 5.5.8.d and 5.5.8.e shall be performed once per 24 months.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

a. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < 0.05% when tested in accordance with ANSl/ASME N510-1989 at the system flowrate specified below:

(continued)

LaSalle 1 and 2 5.5-6 Amendment No. 223/209

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NINE MILE POINT NUCLEAR STATION, LLC EXELON GENERATION COMPANY, LLC DOCKET NO. 50-220 NINE MILE POINT NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 227 Renewed License No. DPR-63

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated July 26, 2016, as supplemented by letter dated October 6, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations 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; 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 amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-63 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, which is attached hereto, as revised through Amendment No. 227, is hereby incorporated into this license. Exelon Generation shall operate the facility in accordance with the Technical Specifications.

Enclosure 12

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION DQJ~a, znch:~

Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: May 2 6 , 2o1 7

ATTACHMENT TO LICENSE AMENDMENT NO. 227 NINE MILE POINT NUCLEAR STATION, UNIT NO. 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-63 DOCKET NO. 50-220 Replace the following pages of the Renewed Facility Operating License and Appendix A, Technical Specifications (TSs), with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert License DPR-63 License DPR-63 Page 3 Page 3 TSs TSs 8 8 108 108 353 353

(2) Exelon Generation pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3) Exelon Generation pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Exelon Generation pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components.

(5) Exelon Generation pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I:

Part 20, Section 30.34 of Part 30; Section 40.41 of Part 40; Section 50.54 and 50.59 of Part 50; and Section 70.32 of Part 70. This renewed license is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect and is also subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 1850 megawatts {thermal).

(2) Technical Specifications The Technical Specifications contained in Appendix A, which is attached hereto, as revised through Amendment No. 227, is hereby incorporated into this license. Exelon Generation shall operate the facility in accordance with the Technical Specifications.

(3) Deleted Renewed License No. DPR-63 Amendment No.191tAFel::J~A21G, 211, 21a, 214, 216, 218, 217, 218, 222, 22a, 224, 226, 227 GeFFeetieR LetteF Dated At:J~t:Jst 7, 2012 CeFFeetieR LetteF DBtee MeFeR 17, 2915 GeFFeetieR LetteF eetee dt:Jly 29, 201 s

1.28 (Deleted) 1.29 (Deleted) 1.30 Reactor Coolant Leakage

a. Identified Leakage (1) Leakage into closed systems, such as pump seal or valve packing leaks that are captured, flow metered and conducted to a sump or collecting tank, or (2) Leakage into the primary containment atmosphere from sources that are both specifically located and known not to be from a through-wall crack in the piping within the reactor coolant pressure boundary.

b. Unidentified Leakage All other leakage of reactor coolant into the primary containment area.

1.31 Core Operating Limits Report The CORE OPERATING LIMITS REPORT is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.6.5. Plant operation within these operating limits is addressed in individual specifications.

1.32 Shutdown Margin (SOM)

SOM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:

a. The reactor is xenon free,

b. The moderator temperature is~ 68° F, corresponding to the most reactive state, and

c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn. With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SOM.

1.33 INSERVICE TESTING PROGRAM The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

AMENDMENT NO. 142. 176. 180. 181~, 227 8

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.7 REACTOR COOLANT SYSTEM ISOLATION VALVES 4.2.7 REACTOR COOLANT SYSTEM ISOLATION VALVES Applicability: Applicability:

Applies to the operating status of the system of Applies to the periodic testing requirement for the isolation valves on lines connected to the reactor reactor coolant system isolation valves.

coolant system.

Objective: Objective:

To assure the capability of the reactor coolant system To assure the capability of the reactor coolant system isolation valves to minimize reactor coolant loss in the isolation valves to minimize reactor coolant loss in the event of a rupture of a line connected to the nuclear event of a rupture of a line connected to the nuclear steam supply system, and to minimize potential leakage steam supply system, and to limit potential leakage paths from the primary containment in the event of a loss- paths from the primary containment in the event of a of-coolant accident. loss-of-coolant accident.

Specification: Specification:

a. Whenever fuel is in the reactor vessel and the reactor The reactor coolant system isolation valves coolant temperature is greater than 212°F, all reactor surveillance shall be performed as indicated below.

coolant system isolation valves on lines connected to the reactor coolant system shall be operable except a. In accordance with the Surveillance as specified in Specification 3.2.7.b below. Frequency Control Program the operable automatically initiated power-operated

b. In the event any isolation valve becomes isolation valves shall be tested for automatic inoperable whenever fuel is in the reactor vessel and initiation and closure times.

the reactor coolant temperature is greater than 212°F, the system shall be considered operable b. Additional surveillances shall be performed as provided that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at least one valve in required by the INSERVICE TESTING PROGRAM.

each line having an inoperable valve is in the mode corresponding to the isolated condition, except as noted in Specification 3.1.1.e.

AMENDMENT N0.142. 145. 173. 181. 197. 222, 227 108

i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives >8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming torn CFR 50, Appendix I;

j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190; and

k. Limitations on venting and purging of the primary containment through the Emergency Ventilation System to maintain releases as low as reasonably achievable.

The provisions of Surveillance Requirements 4.0.2 and 4.0.3 are applicable to the Radioactive Effluent Controls Program surveillance frequencies.

6.5.4 DELETED 6.5.5 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Main Condenser Offgas Treatment System and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

AMENDMENT NO. 142. 157. 162. 181. 182. 199, 227 353

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NINE MILE POINT NUCLEAR STATION, LLC LONG ISLAND LIGHTING COMPANY EXELON GENERATION COMPANY, LLC DOCKET NO. 50-410 NINE MILE POINT NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 161 Renewed License No. NPF-69

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated July 26, 2016, as supplemented by letter dated October 6, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations 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; 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 amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-69 is hereby amended to read as follows:

Enclosure 13

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, as revised through Amendment No. 161, are hereby incorporated into this license.

Exelon Generation shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION c)J 9 tV'---'

David J. Wrona, Branch Chief Plant Licensing Branch 111 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: May 26, 201 7

ATTACHMENT TO LICENSE AMENDMENT NO. 161 NINE MILE POINT NUCLEAR STATION. UNIT NO. 2 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-69 DOCKET NO. 50-410 Replace the following pages of the Renewed Facility Operating License and Appendix A, Technical Specifications (TSs), with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove License NPF-69 License NPF-69 Page4 Page 4 TSs TSs 1.1-3 1.1-3 3.1.7-3 3.1.7-3 3.4.4-1 3.4.4-1 3.4.6-3 3.4.6-3 3.5.1-5 3.5.1-5 3.5.2-4 3.5.2-4 3.6.1.3-11 3.6.1.3-11 3.6.1.3-12 3.6.1.3-12 3.6.2.3-2 3.6.2.3-2 3.6.2.4-2 3.6.2.4-2 5.5-4 5.5-4 5.5-5 5.5-5

(1) Maximum Power Level Exelon Generation is authorized to operate the facility at reactor core power levels not in excess of 3988 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, as revised through Amendment No. 161, are hereby incorporated into this license. Exelon Generation shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Fuel Storage and Handling (Section 9.1, SSER 4)*

a. Fuel assemblies, when stored in their shipping containers, shall be stacked no more than three containers high.

b. When not in the reactor vessel, no more than three fuel assemblies shall be allowed outside of their shipping containers or storage racks in the New Fuel Vault or Spent Fuel Storage Facility.

c. The above three fuel assemblies shall maintain a minimum edge-to-edge spacing of twelve (12) inches from the shipping container array and approved storage rack locations.

d. The New Fuel Storage Vault shall have no more than ten fresh fuel assemblies uncovered at any one time.

(4) Turbine System Maintenance Program (Section 3.5.1.3.10, SER)

The operating licensee shall submit for NRC approval by October 31, 1989, a turbine system maintenance program based on the manufacturer's calculations of missile generation probabilities.

(Submitted by NMPC letter dated October 30, 1989 from C.D. Terry and approved by NRC letter dated March 15, 1990 from Robert Martin to Mr. Lawrence Burkhardt, Ill).

The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report (SER) and/or its supplements wherein the license condition is discussed.

Renewed License No. NPF-69 Amendment 117 tAre1:1~A 140, 141, 143, 144, 146, 146, 147, 168, 161, 162, 164, 166, 167, 168, 169, 169, 161

Definitions 1.1 1.1 Definitions (continued)

EMERGENCY CORE COOLING The ECCS RESPONSE TIME shall be that time interval SYSTEM(ECCS)RESPONSE from when the monitored parameter exceeds its ECCS TIME initiation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

END OF CYCLE The EOC-RPT SYSTEM RESPONSE TIME shall be that RECIRCULATION PUMP TRIP time interval from initial movement of the (EOC-RPT) SYSTEM RESPONSE associated turbine stop valves or turbine control TIME valves to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

INSERVICE TESTING PROGRAM The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

ISOLATION SYSTEM The ISOLATION SYSTEM RESPONSE TIME shall be that RESPONSE TIME time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves travel to their required positions. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE into the drywall such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or (continued)

NMP2 1.1-3 Amendment 91, 125, 161

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.7.7 Verify each pump develops a flow rate In accordance

?. 41.2 gpm at a discharge pressure with the INSERVICE

?, 1335psig. TESTING PROGRAM SR 3.1.7.8 Verify flow through one SLC subsystem In accordance with the from pump into reactor pressure vessel. Surveillance Frequency Control Program SR 3.1.7.9 Verify all heat traced piping between In accordance with the storage tank and pump suction valve is Surveillance Frequency unblocked. Control Program Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after piping temperature is restored to 2!. 70°F SR 3.1.7.10 Verify sodium pentaborate enrichment Prior to is ?. 92 atom percent B-10. addition to SLC tank NMP2 3.1.7-3 Amendment 91, 111, 117, 123, 140, 143, 151, 152, 161

S/RVs 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 Safety/Relief Valves (S/RVs)

LCO 3.4.4 The safety function of 16 S/RVs shall be OPERABLE, APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> S/RVs inoperable.

AND A.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SU RV El LLANCE FREQUENCY SR 3.4.4.1 Verify the safety function lift setpoints In accordance of the required S/RVs are as follows: with the IN SERVICE Number of Setpoint TESTING S/RVs (psig) PROGRAM 2 1165 psig +/- 35.0 4 1175 psig +/- 35.0 4 1185 psig +/- 36.0 4 1195 psig +/- 36.0 4 1205 psig +/- 36.0 Following testing, lift settings shall be within+/- 1%.

NMP2 3.4.4-1 Amendment 9+, 161

RCS PIV Leakage 3.4.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 --------------------------NOTE-------------------------------

Only required to be performed in MODES 1 and2.

Verify equivalent leakage of each RCS PIV In accordance is :s; 0.5 gpm per nominal inch of valve size with the up to a maximum of 5 gpm, at an RCS INSERVICE pressure;:: 1000 psig and ::; 1040 psig. TESTING PROGRAM NMP2 3.4.6-3 Amendment Si-, 161

ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.4 Verify each ECCS pump develops the In accordance specified flow rate with the specified with the developed head. IN SERVICE TESTING TOTAL PROGRAM SYSTEM FLOW RATE DEVELOPED HEAD LPCS ~ 6350 gpm ~ 284 psid LPCS A, B  ;:,>: 7450 gpm  ;:.>: 127 psid LPCIC  ;::: 7450 gpm ~ 140 psid HPCS ~ 6350 gpm  ;::: 327 psid SR 3.5.1.5 -----------------------------NOTE-----------------------------

Vessel injection/spray may be excluded.

Verify each ECCS injection/spray subsystem In accordance with actuates on an actual or simulated the Surveillance automatic initiation signal. Frequency Control Program SR 3.5.1.6 ----------------------------NOTE---------*--------------------

Valve actuation may be excluded.

Verify the ADS actuates on an actual or In accordance with simulated automatic initiation signal. the Surveillance Frequency Control Proa ram SR 3.5.1.7 -----------------------------NOTE------------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify each required ADS valve actuator In accordance with strokes when manually actuated. the Surveillance Frequency Control Program (continued)

NMP2 3.5.1-5 Amendment 91, 152, 161

ECCS - Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2.5 Verify each required ECCS pump develops the In accordance specified flow rate with the specified with the developed head. INSERVICE TESTING TOTAL PROGRAM SYSTEM FLOW RATE DEVELOPED HEAD LPCS  ;:: 6350 gpm  ;::284 psid LPCIA,8  ;:: 7450 gpm  ;:: 127 psid LPCIC  ;:: 7450 gpm  ;:: 140 psid HPCS  ;:: 6350 gpm ~ 327 psid SR 3.5.2.6 ----------------------------NOTE----------------------------------

Vessel injection/spray may be excluded.

Verify each required ECCS injection/spray In accordance with subsystem actuates on an actual or the Surveillance simulated automatic initiation signal. Frequency Control Program SR 3.5.2.7 ---------------------------NOTE----------------------------------

lnstrumentation response time may be assumed to be the design instrumentation response time.

Verify the ECCS RESPONSE TIME for each EGGS In accordance with injection/spray subsystem is within limits. the Surveillance Frequency Control Program NMP2 3.5.2-4 Amendment 91, 152, 161

PC IVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.3 -------------------------- NOTES----------------------------

1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2. Not required to be met for PCIVs that are open under administrative controls.

Verify each primary containment isolation Prior to manual valve and blind flange that is entering MODE 2 located inside primary containment and or 3 from not locked, sealed, or otherwise secured MODE 4, if and is required to be closed during primary accident conditions is closed. containment was de-inerted while in MODE 4, if not performed within the previous 92 days SR 3.6.1.3.4 Verify continuity of the traversing In accordance with incore probe (TIP) shear isolation valve the Surveillance explosive charge. Frequency Control Program SR 3.6.1.3.5 Verify the isolation time of each power In accordance operated, automatic PCIV, except MSIVs, with the is within limits. INSERVICE TESTING PROGRAM (continued)

NMP2 3.6.1.3-11 Amendment 91, 152, 161

PC IVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.6 Perform leakage rate testing for each In accordance with the primary containment purge valve with Surveillance resilient seals. Frequency Control Program AND Once within 92 days after opening the valve SR 3.6.1.3.7 Verify the isolation time of each MSIV is In accordance

~ 3 seconds and ~ 5 seconds. with the INSERVICE TESTING PROGRAM SR 3.6.1.3.B Verify each automatic PCIV actuates to In accordance with the the isolation position on an actual or Surveillance simulated isolation signal. Frequency Control Program SR 3.6.1.3.9 Verify a representative sample of reactor In accordance with the instrumentation line EFCVs actuates to Surveillance the isolation position on an actual or Frequency Control simulated instrument line break signal. Program SR 3.6.1.3.10 Remove and test the explosive squib from In accordance with the each shear isolation valve of the TIP Surveillance System. Frequency Control Program SR 3.6.1.3.11 Verify the leakage rate for the secondary In accordance containment bypass leakage when with 10 CFR 50 pressurized to ;;:: 40 psig is: Appendix J Testing Program

a. Bypass (Drywall): s 8.74 SCFH; and Plan

b. Bypass (Suppression Chamber): s 1.67 SCFH; and

c. Bypass (Drywall with delays): s 28.17 SCFH (continued)

NMP2 3.6.1.3-12 Amendment 91, 96, 152, 156, 161

AHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.3.1 Verify each AHR suppression pool cooling In accordance with subsystem manual and power operated valve the Surveillance in the flow path that is not locked, Frequency Control sealed, or otherwise secured in position, Program is in the correct position or can be aligned to the correct position.

SR 3.6.2.3.2 Verify each required AHA pump develops a In accordance flow rate~ 7450 gpm through the with the associated heat exchanger while operating INSERVICE in the suppression pool cooling mode. TESTING PROGRAM SR 3.6.2.3.3 Verify RHR suppression pool cooling subsystem In accordance with locations susceptible to gas accumulation are the Surveillance sufficiently filled with water. Frequency Control Program NMP2 3.6.2.3-2 Amendment 91, 150, 152, 161

RHR Suppression Pool Spray 3.6.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.4.1 Verify each RHR suppression pool spray In accordance with subsystem manual and power operated valve the Surveillance in the flow path that is not locked, Frequency Control sealed, or otherwise secured in position, Program is in the correct position or can be aligned to the correct position.

SR 3.6.2.4.2 Verify each required AHR pump develops a In accordance flow rate ~ 450 gpm while operating in with the the suppression pool spray mode. INSERVICE TESTING PROGRAM SR 3.6.2.4.3 Verify AHR suppression pool spray subsystem In accordance with locations susceptible to gas accumulation are the Surveillance sufficiently filled with water. Frequency Control Program NMP2 3.6.2.4-2 Amendment91, 150, 152, 161

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program (continued}

1. For noble gases: a dose rate :5 500 mrems/yr to the whole body and a dose rate :5 3000 mrems/yr to the skin, and

2. For iodine-131, iodine-133, tritium, and all radlonuclides in particulate form with half lives greater than 8 days: a dose rate :5 1500 mrems/yr to any organ;

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I;

i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I;

j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190; and

k. Limitations on venting and purging of the primary containment through the Standby Gas Treatment System to maintain releases as low as reasonably achievable.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluent Controls Program surveillance frequencies.

5.5.5 Component Cyclic or Transient Limit This program provides controls to track the USAA, Table 3.98-1 Note 5, cyclic and transient occurrences to ensure that components are maintained within the design limits.

5.5.6 DELETED (continued)

NMP2 5.5-4 Amendment 9-:1-, 161

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP)

I The VFTP shall establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems.

Tests described in Specification 5.5.7.a and 5.5.7.b shall be performed once per 24 months; after each complete or partial replacement of the HEPA filter bank or charcoal adsorber bank; after any structural maintenance on the HEPA filter bank or charcoal adsorber bank housing; and, following significant painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation.

(continued)

NMP2 5.5-5 Amendment 91, 129, 161

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC PSEG NUCLEAR LLC DOCKET NO. 50-277 PEACH BOTTOM ATOMIC POWER STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 313 Renewed License No. DPR-44

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated July 26, 2016, as supplemented by letter dated October 6, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations 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; 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 amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Renewed Facility Operating License No. DPR-44 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 313, are hereby incorporated in the Enclosure 14

license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Q~na~ran~i;-

Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: May 2 6, 2O1 7

ATTACHMENT TO LICENSE AMENDMENT NO. 313 PEACH BOTTOM ATOMIC POWER STATION, UNIT 2 RENEWED FACILITY OPERATING LICENSE NO. DPR-44 DOCKET NO. 50-277 Replace the following pages of the Renewed Facility Operating License and Appendix A, Technical Specifications (TSs), with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove License DPR-44 License DPR-44 Page 3 Page 3 TSs TSs 1.1-3 1.1-3 3.1-23 3.1-23 3.4-9 3.4-9 3.5-5 3.5-5 3.6-15 3.6-15 3.6-28 3.6-28 3.6-39 3.6-39 5.0-11 5.0-11

(5) Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not to separate, such byproduct and special nuclear material as may be produced by operation of the facility, and such Class B and Class C low-level radioactive waste as may be produced by the operation of Limerick Generating Station, Units 1 and 2.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section 70.32 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

(1) Maximum Power Level Exelon Generation Company is authorized to operate the Peach Bottom Atomic Power Station, Unit 2, at steady state reactor core power levels not in excess of 3951 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 313, are hereby incorporated in the license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications.

(3) Physical Protection Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans1, submitted by letter dated May 17, 2006, is entitled: "Peach Bottom Atomic Power Station Security Plan, Training and Qualification Plan, Safeguards Contingency Plan, and Independent Spent Fuel Storage Installation Security Program, Revision 3." The set contains Safeguards Information protected under 10 CFR 73.21.

Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP),

including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Exelon Generation Company CSP was approved by License Amendment No. 281 and modified by Amendment No. 301.

(4) Fire Protection The Exelon Generation Company shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report for the facility, and as approved in the NRC Safety Evaluation Report (SER) dated May 23, 1979, and Supplements dated August 14, September 15, October 10 and November 24, 1980, and in the NRC SERs dated September 16, 1993, and August 24, 1994, subject to the following provision:

1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.

Renewed License No. DPR-44 Revised by letter dated October 28, 2004 Revised by letter dated May 29, 2007 Amendment No. 313 Page 3

Definitions 1.1 1.1 Definitions (continued)

END OF CYCLE The EOC-RPT SYSTEM RESPONSE TIME shall be that RECIRCULATION PUMP TRIP time interval from initial signal generation by (EOC-RPT) SYSTEM RESPONSE the associated turbine stop valve limit switch or TIME from when the turbine control valve hydraulic oil control oil pressure drops below the pressure switch setpoint to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

INSERVICE TESTING PROGRAM The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.SSa(f).

LEAKAGE LEAKAGE shall be:

a. Identified LEA!(AGE

1. LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or

2. LEAKAGE into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;

b. Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE;

c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE;

d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per RATE (LHGR) unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

(continued)

PBAPS UNIT 2 1.1-3 Amendment No. 313

SLC System 3 .1. 7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3 .1. 7. 7 Deleted SR 3.1.7.8 Verify each pump develops a flow rate In accordance

~ 49.1 gpm at a discharge pressure with the

~ 1275 psig. INSERVICE TESTING PROGRAM SR 3.1.7.9 Verify flow through one SLC subsystem from In accordance pump into reactor pressure vessel. with the Surveillance Frequency Control Program.

SR 3.1.7.10 Verify sodium pentaborate enrichment is Prior to

~ 92.0 atom percent B-10. addition to SLC tank PBAPS UNIT 2 3 .1-23 Amendment No. 313

SRVs and SVs 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift setpoints In accordance of the required SRVs and SVs are as with the follows: IN SERVICE TESTING PROGRAM Number of Setpoint SRVs Cpsig) 4 1135 +/- 34.1 4 1145 +/- 34.4 3 1155 +/- 34.7 Number of Setpoint SVs Cpsig) 3 1260 +/- 37.8 Following testing, lift settings shall be within +/- 1%.

SR 3.4.3.2 Verify each required SRV actuator strokes In accordance when manually actuated in the with the depressurization mode. Surveillance Frequency Control Program.

PBAPS UNIT 2 3.4-9 Amendment No. 313

ECCS-Operati ng 3.5.l SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3. 5 .1. 5 Verify each recirculation pump discharge In accordance valve cycles through one complete cycle of with the full travel or is de-energized in the IN SERVICE closed position. TESTING PROGRAM.

SR 3. 5 .1. 6 Verify automatic transfer of the power In accordance supply from the normal source to the with the alternate source for each LPCI subsystem Surveillance inboard injection valve and each Frequency recirculation pump discharge valve. Control Program.

SR 3. 5 .1. 7 -------------------NOTE--------------------

For the core spray pumps, SR 3.5.1.7 may be met using equivalent values for flow rate and test pressure determined using pump curves.

Verify the following ECCS pumps develop the In accordance specified flow rate against a system head with the corresponding to the specified reactor Surveillance pressure. Frequency SYSTEM HEAD Control NO. CORRESPONDING Program.

OF TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF Core Spray  ;::: 3,125 gpm 1  ;::: 105 psig LPCI <! 8,600 gpm 1 <! 20 psig (continued)

PBAPS UNIT 2 3.5-5 Amendment No. 313

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3. 6 .1. 3. 8 Verify the isolation time of each In accordance automatic power operated PCIV, except for with the MSIVs, is within limits. INSERVICE TESTING PROGRAM SR 3. 6 .1. 3. 9 Verify the isolation time of each MSIV is In accordance

~ 3 seconds and ~ 5 seconds. with the INSERVICE TESTING PROGRAM SR 3. 6 .1. 3 .10 Verify each automatic PCIV actuates to In accordance the isolation position on an actual or with the simulated isolation signal. Surveillance Frequency Control Program.

SR 3.6.1.3.11 Verify a representative sample of reactor In accordance instrumentation line EFCVs actuates to with the the isolation position on a simulated Survei 11 ance instrument line break signal. Frequency Control Program.

SR 3. 6 .1. 3 .12 Remove and test the explosive squib from In accordance each shear isolation valve of the TIP with the System. Survei 11 ance Frequency Control Program.

SR 3 . 6 .1. 3. 13 Verify the CAD System supplies nitrogen In accordance to the SGIG System upon loss of the with the normal air supply. Surveillance Frequency Control Program.

(continued)

PBAPS UNIT 2 3.6-15 Amendment No. 313

RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.3.1 Verify each RHR suppression pool cooling In accordance subsystem manual, power operated, and with the automatic valve in the flow path that is Surveillance not locked, sealed, or otherwise secured Frequency in position is in the correct position or Control can be aligned to the correct position. Program.

SR 3.6.2.3.2 Verify each required RHR pump develops a In accordance flow rate ~ 8,600 gpm through the with the associated heat exchanger while operating INSERVICE in the suppression pool cooling mode. TESTING PROGRAM SR 3.6.2.3.3 Verify manual transfer capability of In accordance power supply for the RHR motor-operated with the flow control valve and the RHR cross-tie Surveillance motor-operated valve from the normal Frequency source to the alternate source. Control Program.

SR 3.6.2.3.4 ------------------NOTE-------------------

HPSW system related components are excluded.

Verify RHR suppression pool cooling In accordance subsystem locations susceptible to gas with the accumulation are sufficiently filled with Surveillance water. Frequency Control Program.

PBAPS UNIT 2 3.6-28 Amendment No. 313

SCI Vs 3.6.4.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.2.1 ------------------NOTES------------------

1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2. Not required to be met for SCIVs that are open under administrative controls.

Verify each secondary containment In accordance isolation manual valve and blind flange with the that is not locked, sealed, or otherwise Surveillance secured and is required to be closed Frequency during accident conditions is closed. Control Program.

SR 3.6.4.2.2 Verify the isolation time of each power In accordance operated automatic SCIV is within limits. with the INSERVICE TESTING PROGRAM SR 3.6.4.2.3 Verify each automatic SCIV actuates to In accordance the isolation position on an actual or with the simulated actuation signal. Surveillance Frequency Control Program.

PBAPS UNIT 2 3.6-39 Amendment No. 313

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.6 DELETED 5.5.7 Ventilation Filter Testing Program CVFTP)

The VFTP shall establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems.

Tests described in Specifications 5.5.7.a, 5.5.7.b, and 5.5.7.c shall be performed:

(continued)

PBAPS UNIT 2 5.0-11 Amendment No. 313

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY. LLC PSEG NUCLEAR LLC DOCKET NO. 50-278 PEACH BOTTOM ATOMIC POWER STATION, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 317 Renewed License No. DPR-56

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated July 26, 2016, as supplemented by letter dated October 6, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations 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; 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 amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Renewed Facility Operating License No. DPR-56 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 317, are hereby incorporated in the license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications.

Enclosure 15

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION f2~na.2nch~~

Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: May 2 6 , 2O1 7

ATTACHMENT TO LICENSE AMENDMENT NO. 317 PEACH BOTTOM ATOMIC POWER STATION, UNIT 3 RENEWED FACILITY OPERATING LICENSE NO. DPR-56 DOCKET NO. 50-278 Replace the following pages of the Renewed Facility Operating License and Appendix A, Technical Specifications (TSs), with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert License DPR-56 License DPR-56 Page 3 Page 3 TSs TSs 1.1-3 1.1-3 3.1-23 3.1-23 3.4-9 3.4-9 3.5-5 3.5-5 3.6-15 3.6-15 3.6-28 3.6-28 3.6-39 3.6-39 5.0-11 5.0-11

(5) Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not to separate, such byproduct and special nuclear material as may be produced by operation of the facility, and such Class Band Class Clow-level radioactive waste as may be produced by the operation of Limerick Generating Station, Units 1 and 2.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section 70.32 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

(1) Maximum Power Level Exelon Generation Company is authorized to operate the Peach Bottom Atomic Power Station, Unit No. 3, at steady state reactor core power levels not in excess of 3951 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 317, are hereby incorporated in the license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications.

(3) Physical Protection Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans1, submitted by letter dated May 17, 2006, is entitled: "Peach Bottom Atomic Power Station Security Plan, Training and Qualification Plan, Safeguards Contingency Plan, and Independent Spent Fuel Storage Installation Security Program, Revision 3." The set contains Safeguards Information protected under 10 CFR 73.21.

Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP),

including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Exelon Generation Company CSP was approved by License Amendment No. 283 and modified by Amendment No. 304.

1 The Training and Qualification Plan and Safeguards Contingency Plan and Appendices to the Security Plan.

Renewed License No. DPR-56 Revised by letter dated October 28, 2004 Revised by letter dated November 5, 2004 Revised by letter dated May 29, 2007 Amendment No. 317 Page 3

Definitions 1.1 1.1 Definitions (continued)

END OF CYCLE The EOC-RPT SYSTEM RESPONSE TIME shall be that RECIRCULATION PUMP TRIP time interval from initial signal generation by (EOC-RPT} SYSTEM RESPONSE the associated turbine stop valve limit switch or TIME from when the turbine control valve hydraulic oil control oil pressure drops below the pressure switch setpoint to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

INSERVICE TESTING PROGRAM The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

LEAKAGE LEAKAGE shall bi~:

a. Identified LEAKAGE

1. LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or

2. LEAKAGE into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;

b. Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE;

c. 'rot al LEAKAGE Sum of the identified and unidentified L8AKAGE;

d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per RATE (LHGR) unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

(continued)

PBAPS UNIT 3 1.1-3 Amendment No. 317

SLC System 3 .1. 7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3 .1. 7. 7 Deleted SR 3.1.7.8 Verify each pump develops a flow rate In accordance

~ 49.1 gpm at a discharge pressure with the

~ 1275 psig. INSERVICE TESTING PROGRAM SR 3.1.7.9 Verify flow through one SLC subsystem from In accordance pump into reactor pressure vessel. with the Surveillance Frequency Control Program.

SR 3.1.7.10 Verify sodium pentaborate enrichment is Prior to

~ 92.0 atom percent B-10. addition to SLC tank PBAPS UNIT 3 3.1-23 Amendment No. 317

SRVs and SVs 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift setpoints In accordance of the required SRVs and SVs are as with the follows: INSERVICE TESTING PROGRAM Number of Setpoint SRVs (psi g) 4 1135 +/- 34.1 4 1145 +/- 34.4 3 1155 +/- 34;7 Number of Setpoint SVs Cpsig) 3 1260 +/- 37.8 Following testing, lift settings shall be within +/- 1%.

SR 3.4.3.2 Verify each required SRV actuator strokes In accordance when manually actuated in the with the depressurization mode. Surveillance Frequency Control Program.

PBAPS UNIT 3 3.4-9 Amendment No. 317

ECCS -Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3. 5 .1. 5 Verify each recirculation pump discharge In accordance valve cycles through one complete cycle of with the full travel or is de-energized in the INSERVICE closed position. TESTING PROGRAM.

SR 3. 5 .1. 6 Verify automatic transfer of the power In accordance supply from the normal source to the with the alternate source for each LPCI subsystem Surveillance inboard injection valve and each Frequency recirculation pump discharge valve. Control Program.

SR 3. 5 .1. 7 -------------------NOTE--------------------

For the core spray pumps, SR 3.5.1.7 may be met using equivalent values for flow rate and test pressure determined using pump curves.

Verify the following ECCS pumps develop the In accordance specified flow rate against a system head with the corresponding to the specified reactor Surveillance pressure. Frequency SYSTEM HEAD Control NO. CORRESPONDING Program.

OF TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF Core Spray  ;;:: 3,125 gpm 1 <:: 105 psig LPCI <:: 8,600 gpm 1 <:: 20 psig (continued)

PBAPS UNIT 3 3.5-5 Amendment No. 317

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.8 Verify the isolation time of each In accordance automatic power operated PCIV, except for with the MSIVs, is within limits. INSERVICE TESTING PROGRAM SR 3. 6 .1. 3. 9 Verify the isolation time of each MSIV is In accordance

~ 3 seconds and ~ 5 seconds. with the INSERVICE TESTING PROGRAM SR 3. 6 .1. 3 .10 Verify each automatic PCIV actuates to In accordance the isolation position on an actual or with the simulated isolation signal. Surveillance Frequency Control Program.

SR 3.6.1.3.11 Verify a representative sample of reactor In accordance instrumentation line EFCVs actuates to with the the isolation position on a simulated Surveillance instrument line break signal. Frequency Control Program.

SR 3. 6 .1. 3 .12 Remove and test the explosive squib from In accordance each shear isolation valve of the TIP with the System. Survei 11 ance Frequency Control Program.

SR 3.6.1.3.13 Verify the CAD System supplies nitrogen In accordance to the SGIG System upon loss of the with the normal air supply. Surveillance Frequency Control Program.

(continued)

PBAPS UNIT 3 3.6-15 Amendment No. 317

RHR Suppression Pool Cooling 3 .6.2. 3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.3.1 Verify each RHR suppression pool cooling In accordance subsystem manual, power operated, and with the automatic valve in the flow path that is Surveillance not locked, sealed, or otherwise secured Frequency in position is in the correct position or Control can be aligned to the correct position. Program.

SR 3.6.2.3.2 Verify each required RHR pump develops a In accordance flow rate ~ 8,600 gpm through the with the associated heat exchanger while operating INSERVICE in the suppression pool cooling mode. TESTING PROGRAM SR 3.6.2.3.3 Verify manual transfer capability of In accordance power supply for the RHR motor-operated with the flow control valve and the RHR cross-tie Surveillance motor-operated valve from the normal Frequency source to the alternate source. Control Program.

SR 3.6.2.3.4 ------------------NOTE-------------------

HPSW system related components are excluded.

Verify RHR suppression pool cooling In accordance subsystem locations susceptible to gas with the accumulation are sufficiently filled with Surveillance water. Frequency Control Program.

PBAPS UNIT 3 3.6-28 Amendment No. 317

SCIVs 3.6.4.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.2.1 ------------------NOTES------------------

1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2. Not required to be met for SCIVs that are open under administrative controls.

Verify each secondary containment In accordance isolation manual valve and blind flange with the that is not locked, sealed, or otherwise Surveillance secured and is required to be closed Frequency during accident conditions is closed. Control Program.

SR 3.6.4.2.2 Verify the isolation time of each power In accordance operated automatic SCIV is within limits. with the INSERVICE TESTING PROGRAM SR 3.6.4.2.3 Verify each automatic SCIV actuates to In accordance the isolation position on an actual or with the simulated actuation signal. Surveillance Frequency Control Program.

PBAPS UNIT 3 3.6-39 Amendment No. 317

Programs and Manuals 5.5 5.5 Programs and Manuals (continued)

s. s. 6 DELETED

5. 5. 7 Ventilation Filter Testing Program CVFTP)

The VFTP shall establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems.

Tests described in Specifications 5.5.7.a, 5.5.7.b, and 5.5.7.c shall be performed:

(continued)

PBAPS UNIT 3 5.0-11 Amendment No. 317

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY. LLC AND MIDAMERICAN ENERGY COMPANY DOCKET NO. 50-254 QUAD CITIES NUCLEAR POWER STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 266 Renewed License No. DPR-29

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated July 26, 2016, as supplemented by letter dated October 6, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I;

8. 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; 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 amended by 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-29 is hereby amended to read as follows:

8. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 266, are hereby incorporated into this renewed operating Enclosure 16

license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION OJ 2 ~~

David J. Wrona, Branch Chief Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: May 26, 201 7

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC AND MIDAMERICAN ENERGY COMPANY DOCKET NO. 50-265 QUAD CITIES NUCLEAR POWER STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 261 Renewed License No. DPR-30

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated July 26, 2016, as supplemented by letter dated October 6, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations 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; 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 amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3. B. of Renewed Facility Operating License No. DPR-30 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 261, are hereby incorporated into this renewed operating Enclosure 17

license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION D~~on9.~ra~c~

Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: May 2 6, 2O1 7

ATTACHMENT TO LICENSE AMENDMENT NOS. 266 AND 261 QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 RENEWED FACILITY OPERATING LICENSES NOS. DPR-29 AND DPR-30 DOCKET NOS. 50-254 AND 50-265 Replace the following pages of the Renewed Facility Operating Licenses and Appendix A, Technical Specifications (TSs), with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove License DPR-29 License DPR-29 Page 4 Page 4 License DPR-30 License DPR-30 Page4 Page 4 TSs TSs 1.1-3 1.1-3 3.1.7-3 3.1.7-3 3.4.3-2 3.4.3-2 3.5.1-4 3.5.1-4 3.5.1-5 3.5.1-5 3.5.2-4 3.5.2-4 3.6.1.3-7 3.6.1.3-7 3.6.2.3-2 3.6.2.3-2 5.5-4 5.5-4 5.5-5 5.5-5

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 266, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

C. The licensee shall maintain the commitments made in response to the March 14, 1983, NUREG-0737 Order, subject to the following provision:

The licensee may make changes to commitments made in response to the March 14, 1983, NUREG-0737 Order without prior approval of the Commission as long as the change would be permitted without NRC approval, pursuant to the requirements of 10 CFR 50.59. Consistent with this regulation, if the change results in an Unreviewed Safety Question, a license amendment shall be submitted to the NRC staff for review and approval prior to implementation of the change.

D. Equalizer Valve Restriction Three of the four valves in the equalizer piping between the recirculation loops shall be closed at all times during reactor operation with one bypass valve open to allow for thermal expansion of water.

E. The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined sets of plans 1 , which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Quad Cities Nuclear Power Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 2," submitted by letter dated May 17, 2006.

Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP}, including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).

The Exelon Generation Company CSP was approved by License Amendment No. 249 as modified by License Amendment No. 259.

F. The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report for the facility and as approved in the Safety Evaluation Reports dated July 27, 1979 with supplements dated November 5, 1980, and 1

The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.

Renewed License No. DPR-29 Amendment No. 266

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 261, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

C. The license shall maintain the commitments made in response to the March 14, 1983, NUREG-0737 Order, subject to the following provision:

The licensee may make changes to commitments made in response to the March 14, 1983, NUREG-0737 Order without prior approval of the Commission as long as the change would be permitted without NRC approval, pursuant to the requirements of 10 CFR 50.59. Consistent with this regulation, if the change results in an Unreviewed Safety Question, a license amendment shall be submitted to the NRC staff for review and approval prior to implementation of the change.

D. Equalizer Valve Restriction Three of the four valves in the equalizer piping between the recirculation loops shall be closed at all times during reactor operation with one bypass valve open to allow for thermal expansion of water.

E. The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans 1 , which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Quad Cities Nuclear Power Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 2," submitted by letter dated May 17, 2006.

Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP),

including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Exelon Generation Company CSP was approved by License Amendment No. 244 and modified by License Amendment No. 254.

F. The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report for the facility and as approved in the Safety Evaluation Reports dated July 27, 1979 with supplements dated 1

The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.

Renewed License No. DPR-30 Amendment No. 261

Defi nit i ans 1.1 1.1 Definitions DOSE EQUIVALENT I-131 Guidance Report 11, "Limiting Values of (continued) Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989.

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE into the drywel 1, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or

2. LEAKAGE into the drywel l atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;

b. Unidentified lEAKAGE A11 LEAKAGE into the drywel 1 that is not identified LEAKAGE;

c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE; and

d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe 1vall, or vessel wall.

(continued)

Quad Cities l and 2 1. 1- 3 Amendment No. 266/261

SLC System 3 .1. 7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.5 Verify the concentration of sodium In accordance pentaborate in solution is within the with the limits of Figure 3.1.7-1. Surveillance Frequency Control Program Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after water or sodium pentaborate is added to solution Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after solution temperature is restored within the limits of Figure 3.1.7-2 SR 3.1.7.6 Verify each SLC subsystem manual valve in In accordance the flow path that is not locked, sealed, with the or otherwise secured in position is in the Surveillance correct position, or can be aligned to the Frequency correct position. Control Program SR 3.1.7.7 Verify each pump develops a flow rate In accordance

~ 40 gpm at a discharge pressure with the

~ 1275 psig. INSERVICE TESTING PROGRAM SR 3.1.7.8 Verify flow through one SLC subsystem from In accordance pump into reactor pressure vessel. with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.1.7-3 Amendment No. 266/261

Safety and Relief Valves 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift setpoints In accordance of the safety valves are as follows: with the INSERVICE Number of Setpoint TESTING PROGRAM Safety Valves (psjg) 1 1135 +/- 34.1 2 1240 +/- 37.2 2 1250 +/- 37.5 4 1260 +/- 37.8 Following testing, lift settings shall be within +/- 1%.

SR 3.4.3.2 Verify each relief valve actuator strokes In accordance when manually actuated. with the Surveillance Frequency Control Program SR 3.4.3.3 - - - - - - - - - - - - - - - - - - - NOTE - - - - - - - - - - - - - - - - - - - -

Valve actuation may be excluded.

Verify each relief valve actuates on an In accordance actual or simulated automatic initiation with the signal. Surveillance Frequency Control Program Quad Cities 1 and 2 3.4.3-2 Amendment No. 266/261

ECCS-Operating

3. 5 .1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify, for each ECCS injection/spray In accordance subsystem, locations susceptible to gas with the accumulation are sufficiently filled with Surveillance water. Frequency Control Program SR 3.5.1.2 - - - - - - - - - - - - - - - - - - - NOTES- - - - - - - - - - - - - - - - - - -

1. Low pressure coolant injection (LPCil subsystems may be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the Residual Heat Removal (RHR) cut-in permissive pressure in MODE 3, if capable of being manually realigned and not otherwise inoperable.

2. Not required to be met for system vent flow paths opened under administrative control.

Verify each ECCS injection/spray subsystem In accordance manual. power operated, and automatic valve with the in the flow path, that is not locked, Surveillance sealed, or otherwise secured in position, Frequency is in the correct position. Control Program SR 3.5.1.3 Verify correct breaker alignment to the In accordance LPCI swing bus. \vi th the Surveillance Frequency Control Program SR 3.5.1.4 Verify each recirculation pump discharge In accordance valve cycles through one complete cycle of with the full travel or is de-energized in the INSERVICE closed position. TES TI NG PROGRAM (continued)

Quad Cities 1 and 2 3.5.1-4 Amendment No. 266/261

ECCS-Operating 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.5 Verify the following ECCS pumps develop the In accordance specified flow rate against a test line with the pressure corresponding to the specified INSERVICE reactor pressure. TESTING PROGRAM TEST LINE PRESSURE NO. CORRESPONDING OF TO A REACTOR SYSTEM FLOW RATE EU11E.5. PRESSURE OF Core Spray ~ 4500 gprn 1 ~ 90 psig LPC I ~ 9000 gpm 2 ~ 20 psig SR 3.5.1.6 - - - - - - - - - - - - - - - - - - - NOTE - - - - - - - - - - - - - - - - - - - -

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure ~ 1005 and In accordance

~ 920 psig, the HPCI pump can develop a with the flow rate~ 5000 gpm against a system head INSERVICE corresponding to reactor pressure. TESTING PROGRAM SR 3.5.1.7 - - - - - - - - - - - - - - - - - - - NOTE - - - - - - - - - - - - - - - - - - - -

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure~ 180 psig, In accordance the HPCI pump can develop a flow rate with the

~ 5000 gpm against a system head Surveillance corresponding to reactor pressure. Frequency Control Program (continued)

Quad Cities 1 and 2 3.5.1-5 Amendment No. 266/261

ECCS-Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.4 Verify each required ECCS pump develops the In accordance specified flow rate against a test line with the pressure corresponding to the specified INSERVICE reactor pressure. TESTING PROGRAM TEST LINE PRESSURE NO. CORRESPONDING OF TO A REACTOR SYSTEM FLOW RATE .PJJJ.1E.S PRESSURE OF cs ~ 4500 gpm 1 ~ 90 psig LPCI  :::: 4500 gpm 1  :::: 20 psig SR 3.5.2.5 - - - - - - - - - - - - - - - - -- -NOTE - - - - --- - -- - -- -- - - - - -

Vessel injection/spray may be excluded.

Verify each required ECCS injection/spray In accordance subsystem actuates on an actual or with the simulated automatic initiation signal. Surveillance Frequency Control Program Quad Cities 1 and 2 3.5.2-4 Amendment No. 266/261

PC I Vs 3.6.1.3 SURVEILLANCE REQUIREMENTS SU RV EI LLANCE FREQUENCY SR 3.6.1.3.4 Verify continuity of the traversing In accordance incore probe (TIP) shear isolation valve with the explosive charge. Surveillance Frequency Control Program SR 3.6.1.3.5 Verify the isolation time of each power In accordance operated, automatic PCIV, except for with the MSIVs, is within limits. INSERVICE TESTING PROGRAM SR 3.6.1.3.6 Verify the isolation time of each MSIV is In accordance

~ 3 seconds and ~ 5 seconds. with the INSERVICE TESTING PROGRAM SR 3.6.1.3.7 Verify each automatic PCIV actuates to In accordance the isolation position on an actual or with the simulated isolation signal. Surveillance Frequency Control Program SR 3.6.1.3.8 Verify a representative sample of reactor In accordance instrumentation line EFCVs actuate to the with the isolation position on an actual or Surveillance simulated instrument line break signal. Frequency Control Program SR 3.6.1.3.9 Remove and test the explosive squib from In accordance each shear isolation valve of the TIP with the System. Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.6.1.3-7 Amendment No. 266/261

RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.3.1 Verify each RHR suppression pool cooling In accordance subsystem manual and power operated valve with the in the flow path that is not locked, Surveillance sealed, or otherwise secured in position, Frequency is in the correct position or can be Control Program aligned to the correct position.

SR 3.6.2.3.2 Verify each required RHR pump develops a In accordance flow rate~ 5000 gpm through the with the associated heat exchanger while operating INS ERV ICE in the suppression pool cooling mode. TESTING PROGRAM SR 3.6.2.3.3 Verify RHR suppression pool cooling In accordance subsystem locations susceptible to gas with the accumulation are sufficiently filled with Surveillance water. Frequency Control Program Quad Cities 1 and 2 3.6.2.3*2 Amendment No. 266/261

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program (continued)

1. For noble gases: a dose rates 500 mrems/yr to the whole body and a dose rate~ 3000 mrems/yr to the skin, and

2. For iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days: a dose rates 1500 mrems/yr to any organ;

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I;

i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives> 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and

j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluents Control Program Surveillance Frequencies.

5.5.5 Component Cyclic or Transient Limit This program provides controls to track the UFSAR Section 3.9, cyclic and transient occurrences to ensure that components are maintained within the design limits.

5.5.6 DELETED (continued)

Quad Cities 1 and 2 5.5-4 Amendment No. 266/261

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP)

The VFTP shall establish the required testing of Engineered Safety Feature (ESE) filter ventilation systems. Tests described in Specification 5.5.7.a and 5.5.7.b shall be performed once per 24 months; after each complete or partial replacement of the HEPA filter bank or charcoal adsorber bank; after any structural maintenance on the HEPA filter bank or charcoal adsorber bank housing; and, following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation that could adversely affect the filter bank or charcoal adsorber capability.

(continued)

Quad Cities 1 and 2 5.5-5 Amendment No. 266/261

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 RE. GINNA NUCLEAR POWER PLANT. LLC EXELON GENERATION COMPANY. LLC DOCKET NO. 50-244 RE. GINNA NUCLEAR POWER PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 124 Renewed License No. DPR-18

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated July 26, 2016, as supplemented by letter dated October 6, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations 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; 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 amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-18 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 124, are hereby incorporated in the renewed license. Exelon Generation shall operate the facility in accordance with the Technical Specifications.

Enclosure 18

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

~~on9.~ra::::hief Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: May 26, 2017

ATTACHMENT TO LICENSE AMENDMENTS NO. 124 R.E. GINNA NUCLEAR POWER PLANT RENEWED FACILITY OPERATING LICENSE NO. DPR-18 DOCKET NO. 50-244 Replace the following pages of the Renewed Facility Operating License and Appendix A, Technical Specifications (TSs), with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove License DPR-18 License DPR-18 Page 3 Page 3 TSs TSs 1.1-2 1.1-2 3.4.10-2 3.4.10-2 3.5.2-3 3.5.2-3 3.6.3-6 3.6.3-6 3.6.6-2 3.6.6-2 3.7.1-2 3.7.1-2 3.7.2-2 3.7.2-2 3.7.3-2 3.7.3-2 3.7.5-3 3.7.5-3 3.7.7-2 3.7.7-2 5.5-4 5.5-4

(b) Exelon Generation pursuant to the Act and 10 CFR Part 70, to possess and use four (4) mixed oxide fuel assemblies in accordance with the RG&E's application dated December 14, 1979 (transmitted by letter dated December 20, 1979), as supplemented February 20, 1980, and March 5, 1980; (3) Exelon Generation pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Exelon Generation pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Exelon Generation pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

(1) Maximum Power Level Exelon Generation is authorized to operate the facility at steady-state power levels up to a maximum of 1775 megawatts (thermal).

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 124, are hereby incorporated in the renewed license.

Exelon Generation shall operate the facility in accordance with the Technical Specifications.

(3) Fire Protection Exelon Generation shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee's amendment request dated March 28, 2013, supplemented by letters dated December 17, 2013; January 29, 2014; February 28, 2014; September 5, 2014; September 24, 2014; December 4, 2014; March 18, 2015; June 11, 2015; August 7, 2015; and as approved in the safety evaluation report dated November 23, 2015. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no R. E. Ginna Nuclear Power Plant Amendment No. 124 I

Definitions 1.1 CHANNEL A COT shall be the injection of a simulated or actual signal into the OPERATIONAL channel as close to the sensor as practicable to verify the OPERABILITY TEST of required alarm, interlock, display, and trip functions. The COT shall (COT) include adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy.

CORE CORE ALTERATIONS shall be the movement of any fuel, sources, or ALTERATIONS reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING The COLA is the plant specific document that provides cycle specific LIMITS REPORT parameter limits for the current reload cycle. These cycle specific (COLA) parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.

DOSE DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 EQUIVALENT 1-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in ICAP 30, Supplement to Part 1, pages 192-212, table entitled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity."

DOSE DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 EQUIVALENT (microcuries per gram} that alone would produce the same acute dose to XE-133 the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, "External Exposure to Radionuclides in Air, Water, and Soil," 1993.

INSERVICE The INSERVICE TESTING PROGRAM is the licensee program that fulfills TESTING the requirements of 10 CFR 50.55a(f).

PROGRAM R.E. Ginna Nuclear Power Plant i.1-2 Amendment m, 124

Pressurizer Safety Valves 3.4.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 -NOTE-Required to be performed within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> of entering MODE 4 from MODE 5 with all RCS cold leg temperatures greater than the LTOP enable temperature specified in the PTLR for the purpose of setting the pressurizer safety valves under ambient (hot) conditions only provided a preliminary cold setting was made prior to heatup.

Verify each pressurizer safety valve is OPERABLE in In accordance with accordance with the INSERVICE TESTING the INSERVICE PROGRAM. Following testing, lift settings shall be TESTING within+/- 1%. PROGRAM R.E. Ginna Nuclear Power Plant 3.4.10-2 Amendment a+, 124

ECCS - MODES 1, 2, and 3 3.5.2 SURVEILLANCE FREQUENCY Verify each breaker or key switch, as applicable, for each In accordance SR3.5.2.3 valve listed in SR 3.5.2.1, is in the correct position. with the Surveillance Frequency Control Program SR 3.5.2.4 Verify each ECCS pump's developed head at the test flow In accordance point is greater than or equal to the required developed with the head. INSERVICE TESTING PROGRAM Verify each ECCS automatic valve in the flow path that is In accordance SR 3.5.2.5 with the not locked, sealed, or otherwise secured in position Surveillance actuates to the correct position on an actual or simulated Frequency actuation signal.

Control Program In accordance SR3.5.2.6 Verify each ECCS pump starts automatically on an actual With the or simulated actuation signal.

!Surveillance Frequency Control Program Verify, by visual inspection, each RHR containment sump In accordance SR3.5.2.7 with the suction inlet is not restricted by debris and the Surveillance containment sump screen shows no evidence of Frequency structural distress or abnormal corrosion.

Control Program SR 3.5.2.8 Verify ECCS locations susceptible to gas accumulation are In accordance sufficiently filled with water. With the Surveillance Frequency Control Program R.E. Ginna Nuclear Power Plant 3.5.2-3 Amendment No.~. 124

Containment Isolation Boundaries 3.6.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify each mini-purge valve is closed, except when In accordance with SR3.6.3.1 the penetration flowpath(s) are permitted to be open lthe Surveillance Frequency Control under administrative control.

Proa ram SR3.6.3.2 .. .,, - ---- - -----NOTE- - ---- ---- - ---- - ----

1. Isolation boundaries in high radiation areas may be verified by use of administrative controls.

2. Not applicable to containment isolation boundaries which receive an automatic containment isolation signal.

Verify each containment isolation boundary that is In accordance with located outside containment and not locked, sealed, he Surveillance or otherwise secured in the required position is Frequency Control performing its containment isolation accident function Program except for containment isolation boundaries that are open under administrative controls .

SR3.6.3.3 ....... - ---- ------NOTE- - ---- ---- - ---- - ----

1. Isolation boundaries in high radiation areas may be verified by use of administrative means.

2. Not applicable to containment isolation boundaries which receive an automatic containment isolation signal.

..~ - -- -- - - - -- - - - -- - - - - - - --- - --- -

Verify each containment isolation boundary that is Prior to entering located inside containment and not locked, sealed, or MODE4from otherwise secured in the required position is MODE 5 if not performing its containment isolation accident function, performed within the except for containment isolation boundaries that are previous 92 days open under administrative controls.

SR 3.6.3.4 Verify the isolation time of each automatic In accordance with containment isolation valve is within limits. the INSERVICE TESTING PROGRAM SR3.6.3.5 Perform required leakage rate testing of containment In accordance with mini-purge valves with resilient seals in accordance the Containment with the Containment Leakage Rate Testing Program. Leakage Rate Program.

R.E. Ginna Nuclear Power Plant 3.6.3-6 Amendment No. +2a, 124

CS, CRFC, and NaOH Systems 3.6.6 CONDITION REQUIRED ACTION COMPLETION TIME F. Two CS trains inoperable. F.1 Enter LCO 3.0.3. Immediately Three or more CRFC units inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1 Perform SR 3.5.2.1 and SR 3.5.2.3 for valves 896A In accordance with and 8968. applicable SRs.

• - - - - - - - - - - - * - - - - -NOTE- - - - - - - - - - * - - - - - - In accordance with SR3.6.6.2 Not required to be met for system vent flow paths he Surveillance opened under administrative control. Frequency Control

• -------------------------------------- Program Verify each CS manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position.

In accordance with SR 3.6.6.3 Verify each NaOH System manual, power operated, the Surveillance and automatic valve in the flow path that is not locked, Frequency Control sealed, or otherwise secured in position is in the Program correct position.

In accordance with SR 3.6.6.4 Operate each CRFC unit for~ 15 minutes.

he Surveillance Frequency Control Program In accordance with SR 3.6.6.5 Verify cooling water flow through each CRFC unit.

lthe Surveillance Frequency Control Program SR 3.6.6.6 Verify each CS pump's developed head at the flow In accordance with test point is greater than or equal to the required the INSERVICE developed head. TESTING PROGRAM In accordance with SR 3.6.6.7 Verify NaOH System solution volume is~ 3000 gal.

he Surveillance Frequency Control Program In accordance with SR3.6.6.8 Verify NaOH System tank NaOH solution lthe Surveillance concentration is ~ 30% and ::;; 35% by weight.

Frequency Control Program R.E. Ginna Nuclear Power Plant 3.6.6-2 Amendment No. .+22, 124

MSSVs 3.7.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.1.1 -NOTE-Only required to be performed in MODES 1 and 2.

Verify each MSSV lift setpoint specified below in In accordance accordance with the INSERVICE TESTING PROGRAM. with the Following testing, lift settings shall be within +/- 1%. INSERVICE TESTING PROGRAM VALVE NUMBER LIFT SETTING SGB (psig +1%. -3%)

3509 3508 1140 3511 3510 1140 3515 3512 1140 3513 3514 1085 R.E. Ginna Nuclear Power Plant 3.7.1-2 Amendment 89, 124

MSIVs and Non-Return Check Valves 3.7.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify closure time of each MSIV is :s; 5 seconds under In accordance with no flow and no load conditions. the INSERVICE TESTING PROGRAM SR3.7.2.2 Verify each main steam non-return check valve can In accordance with close. the INSERVICE TESTING PROGRAM In accordance with SR3.7.2.3 Verify each MSIV can close on an actual or simulated

~he Surveillance actuation signal.

Frequency Control Program A.E. Ginna Nuclear Power Plant 3.7.2-2 Amendment No.~. 124

MFIVs, MFRVs, and Associated Bypass Valves 3.7.3 CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and E.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met.

E.2 Bein MODE4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1 Verify the closure time of each MFIV is :s; 30 seconds In accordance with on an actual or simulated actuation signal. the INSERVICE TESTING PROGRAM SR 3.7.3.2 Verify the closure time of each MFRV and associated In accordance with bypass valve is ~ 1Oseconds on an actual or the INSERVICE simulated actuation signal. TESTING PROGRAM R.E. Ginna Nuclear Power Plant 3.7.3-2 Amendment~. 124

AFW System 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 Verify each AFW and SAFW manual, power operated, In accordance with and automatic valve in each water flow path, and in he Surveillance both steam supply flow paths to the turbine driven Frequency Control pump, that is not locked, sealed, or otherwise secured Program in position, is in the correct position.

SR 3.7.5.2 *--------------------------------*

-NOTE-Required to be met prior to entering MODE 1 for the TDAFWpump.

Verify the developed head of each AFW pump at the In accordance with flow test point is greater than or equal to the required the INSERVICE developed head. TESTING PROGRAM SR3.7.5.3 Verify the developed head of each SAFW pump at the In accordance with I

flow test point is greater than or equal to the required the INSERVlCE developed head. TESTING PROGRAM SR 3.7.5.4 Perform a complete cycle of each AFW and SAFW In accordance with motor operated suction valve from the Service Water the INSERVICE System, each AFW and SAFW discharge motor TESTING operated isolation valve, and each SAFW cross-tie PROGRAM motor operated valve.

In accordance with SR 3.7.5.5 Verify each AFW automatic valve that is not locked,

~he Surveillance sealed, or otherwise secured in position, actuates to Frequency Control the correct position on an actual or simulated Program actuation signal.

SR 3.7.5.6 ---------------------------------*

-NOTE-Required to be met prior to entering MODE 1 for the TDAFWpump.

Verify each AFW pump starts automatically on an In accordance with actual or simulated actuation signal. the Surveillance Frequency Control Program In accordance with SR3.7.5.7 Verify each SAFW train can be actuated and

~he Surveillance controlled from the control room.

Frequency Control Program R.E. Ginna Nuclear Power Plant 3.7.5-3 Amendment No.~. 124

CCW System 3.7.7 CONDITION REQUIRED ACTION COMPLETION TIME D.2 Bein MODE3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> D.3 Be in MODE4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.7.1 -NOTE-Isolation of CCW flow to individual components does not render the CCW loop header inoperable.

Verify each CCW manual and power operated valve In accordance with in the CCW train and heat exchanger flow path and the Surveillance loop header that is not locked, sealed, or otherwise Frequency Control secured in position, is in the correct position. Program SR3.7.7.2 Perform a complete cycle of each motor operated In accordance with isolation valve to the residual heat removal heat the INSERVICE exchangers. TESTING PROGRAM R.E. Ginna Nuclear Power Plant 3.7.7-2 Amendment No.~. 124

Programs and Manuals 5.5 5.5.7 DELETED R.E. Ginna Nuclear Power Plant 5.5-4 Amendment +w, 124

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY. LLC DOCKET NO. 50-289 THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 290 Renewed License No. DPR-50

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated July 26, 2016, as supplemented by letter dated October 6, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations 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; 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 amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.c.(2) of Renewed Facility Operating License No. DPR-50 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 290, are hereby incorporated in the license. The Exelon Generation Company shall operate the facility in accordance with the Technical Specifications.

Enclosure 19

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

~~09.*. Br~~ef Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: May 2 6 , 2 O1 7

ATTACHMENT TO LICENSE AMENDMENT NO. 290 THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289 Replace the following pages of the Renewed Facility Operating License and Appendix A, Technical Specifications (TSs), with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove License DPR-50 License DPR-50 Page 4 Page 4 TSs TSs 1-8 1-8 4-8 4-8 4-11 4-11 4-52 4-52

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 290 , are hereby incorporated in the license. The Exelon Generation Company shall operate the facility in accordance with the Technical Specifications.

(3) Physical Protection Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p ). The combined set of plans 1, submitted by letter dated May 17, 2006, is entitled: "Three Mile Island Nuclear Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 3." The set contains Safeguards Information protected under 10 CFR 73.21.

Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Exelon Generation Company CSP was approved by License Amendment No. 275 and modified by License Amendment No. 288.

(4) Fire Protection Exelon Generation Company shall implement and maintain in effect all provisions of the Fire Protection Program as described in the Updated FSAR for TMl-1.

Changes may be made to the Fire Protection Program without prior approval by the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. Temporary changes to specific fire protection features which may be necessary to accomplish maintenance or modifications are acceptable provided that interim compensate measures are implemented.

(5) The licensee shall implement a secondary water chemistry monitoring program to inhibit steam generator tube degradation. This program shall include:

a. Identification of a sampling schedule for the critical parameters and control points for these parameters;

b. Identification of the procedures used to measure the values of the critical parameters;

c. Identification of process sampling points;

d. Procedure for the recording and management of data; 1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.

Amendment No. 290 Renewed Operating License No. DPR-50

1.24 CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT is a TMI-1 specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.5. Plant operation within these operating limits is addressed in individual specifications.

1.25 FREQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2. All Surveillance Requirements shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the surveillance interval. The 25% extension applies to all frequency intervals with the exception of "F." No extension is allowed for intervals designated "F."

TABLE 1.2 FREQUENCY NOTATION NOTATION FREQUENCY s Shiftly (once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />)

D Daily (once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) w Weekly (once per 7 days)

M Monthly (once per 31 days)

Q Quarterly (once per 92 days)

SIA Semi-Annually (once per 184 days)

R Refueling Interval (once per 24 months)

PS/U Prior to each reactor startup, if not done during the previous 7 days PS/A Within six (6) months prior to each reactor startup p Completed prior to each release N/A(NA) Not applicable E Once per 18 months F Not to exceed 24 months 1.26 DOSE EQUIVALENT Xe-133 Dose Equivalent Xe-133 shall be that concentration ofXe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-13lm, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT Xe-133 shall be performed using effective dose conversion factors for air submersion listed in Table III. l of EPA Federal Guidance Report No. 12.

l .27 INSERVICE TESTING PROGRAM The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

1-8 Amendment No. 72, 137, 155, 173, 175, 199, 272, 290

TABLE 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY Item Frequency

1. Control Rods Rod drop times of all Note 1 full length rods

2. Control Rod Movement of each rod Note 1, when reactor is Movement critical

3. Pressurizer Setpoint In accordance with the Safety Valves INSERVICE TESTING PROGRAM

4. Main Steam Setpoint In accordance with the Safety Valves INSERVICE TESTING PROGRAM

5. Refueling System Functional Start of each Interlocks refueling period

6. (Deleted)

7. Reactor Coolant Evaluate Note 1, when reactor System Leakage coolant system temperature is greater than 525 degrees F (Not applicable to primary-to-secondary leakage.)

8. {Deleted)

9. Spent Fuel Functional Each refueling period Cooling System prior to fuel handling

10. Intake Pump (a) Silt Accumulation - Note 1 House Floor Visual inspection (Elevation of Intake Pump 262 ft. 6 in.) House Floor (b) Silt Accumulation Note 1 Measurement of Pump House Flow

11. Pressurizer Block Functional* Note 1 Valve (RC-V2)

12. Primary to Secondary Evaluate Note 1 (Note: Not required Leakage to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.)

• Function shall be demonstrated by operating the valve through one complete cycle of full travel.

Note 1: Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

4-8 Amendment No. 5&, 68, 78, ~ * .:l-7a, :j..g.S, ~. 246, 261, 274, 290

4.2 REACTOR COOLANT SYSTEM INSERVICE AND TESTING Applicability This technical specification applies to the inservice inspection (ISi) of the reactor coolant system pressure boundary and portions of other safety oriented system pressure boundaries.

Objective The objective of the ISi program is to provide assurance of the continuing integrity of the reactor coolant system while at the same time minimizing radiation exposure to personnel in the performance of inservice inspections.

Specification 4.2.1 ISi of ASME Code Class 1, Class 2, and Class 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a, except where specific written relief has been granted by the NRG.

4.2.2 DELETED.

4.2.3 (Deleted) 4.2.4 The accessible portions of one reactor coolant pump motor flywheel assembly will be ultrasonically inspected within the first ISi period, two reactor coolant pump motor flywheel assemblies within the first two ISi periods and all four by the end of the 1O year inspection interval. However, the U.T. procedure is developmental and will be used only to the extend that it is shown to be meaningful. The extent of coverage will be limited to those areas of the flywheel which are accessible without motor disassembly, i.e., can be reached through the access ports. Also, if radiation levels at the lower access ports are prohibitive, only the upper access ports will be used.

4-11 Amendment No. 15, 29, 54, 60, 71, ii8, 172, 266, 290

4.9 DECAY HEAT REMOVAL (OHR) CAPABILITY- PERIODIC TESTING Applicability Applies to the periodic testing of systems or components which function to remove decay heat.

Objective To verify that systems/components required for DHR are capable of performing their design function.

Specification 4.9.1 Reactor Coolant System (RCS) Temperature greater than 250 degrees F.

4.9.1.1 Verify each Emergency Feedwater (EFW) Pump is tested in accordance with the requirements and acceptance criteria of the INSERVICE TESTING PROGRAM.

Note: This surveillance is not required to be performed for the turbine-driven EFW Pump (EF-P-1) until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding 750 psig.

4.9.1.2 DELETED 4.9.1.3 At the frequency specified in the Surveillance Frequency Control Pegram, each EFW System flowpath valve from both Condensate Storage Tanks (CSTs) to the OTSGs via the motor-driven pumps and the turbine-driven pump shall be verified to be in the required status.

4.9.1.4 At the frequency specified in the Surveillance Frequency Control Program:

a) Verify that each EFW Pump starts automatically upon receipt of an EFW test signal.

b) Verify that each EFW control valve responds upon receipt of an EFW test signal.

c) Verify that each EFW control valve responds in manual control from the control room and remote shutdown panel.

4.9.1.5 Prior to STARTUP, following a REFUELING SHUTDOWN or a COLD SHUTDOWN greater than 30 days, conduct a test to demonstrate that the motor driven EFW Pumps can pump water from the CSTs to the Steam Generators.

4-52 Amendment No. 78, 119, 124, 172, 242, 266, 274, 290

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 191 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-72, AMENDMENT NO. 192 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-77, AMENDMENT NO. 197 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-37, AMENDMENT NO. 197 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-66, AMENDMENT NO. 320 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-53.

AMENDMENT NO. 298 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-69, AMENDMENT NO. 212 TO FACILITY OPERATING LICENSE NO. NPF-62, AMENDMENT NO. 254 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-19, AMENDMENT NO. 247 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-25, AMENDMENT NO. 223 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-11.

AMENDMENT NO. 209 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-18, AMENDMENT NO. 161 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-69, AMENDMENT NO. 313 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-44, AMENDMENT NO. 317 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-56, AMENDMENT NO. 266 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-29, AMENDMENT NO. 261 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-30, AND AMENDMENT NO. 124 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18.

EXELON GENERATION COMPANY, LLC BRAIDWOOD STATION, UNITS 1 AND 2 BYRON STATION, UNIT NOS. 1 AND 2 CALVERT CLIFFS NUCLEAR POWER PLANT. UNITS 1 AND 2 Enclosure 20

CLINTON POWER STATION, UNIT NO. 1 DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 LASALLE COUNTY STATION, UNITS 1 AND 2 NINE MILE POINT NUCLEAR STATION, UNIT 2 PEACH BOTTOM ATOMIC POWER STATION, UNITS 2 AND 3 QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 R. E. GINNA NUCLEAR POWER PLANT DOCKET NOS. STN 50-456, STN 50-457, STN 50-454, STN 50-455, 50-317, 50-318, 50-461, 50-237, 50-249, 50-373, 50-374, 50-410, 50-277, 50-278, 50-254, 50-265, AND 50-244

1.0 INTRODUCTION

By application dated July 26, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16209A218), as supplemented by letter dated October 6, 2016 (ADAMS Accession No. ML16280A402), Exelon Generation Company, LLC (the licensee) requested changes to the technical specifications (TSs) for Braidwood Station (Braidwood),

Units 1 and 2; Byron Station (Byron), Unit Nos. 1 and 2; Calvert Cliffs Nuclear Power Plant (Calvert Cliffs), Units 1 and 2; Clinton Power Station (Clinton), Unit No. 1; Dresden Nuclear Power Station (Dresden), Units 2 and 3; LaSalle County Station (LaSalle), Units 1 and 2; Nine Mile Point Nuclear Station (Nine Mile Point), Unit 2; Peach Bottom Atomic Power Station (Peach Bottom), Units 2 and 3; Quad Cities Nuclear Power Station (Quad Cities), Units 1 and 2; and R.

E. Ginna Nuclear Power Plant (Ginna) (the facilities). The licensee's October 6, 2016, letter revised the scope of the TS changes requested and withdrew the proposed TS changes originally requested in the July 26, 2016, application.

Specifically, the licensee requested changes to the TSs consistent with Technical Specifications Task Force (TSTF) Standard Technical Specifications (STS) Change Traveler TSTF-545, Revision 3, "TS lnservice Testing Program Removal & Clarify SR [Surveillance Requirement]

Usage Rule Application to Section 5.5 Testing," dated October 21, 2015 (ADAMS Accession No. ML15294A555). For each facility, the licensee's proposed changes delete the lnservice Testing Program from TS Section 5.5, "Programs and Manuals," and adds a new defined term, "INSERVICE TESTING PROGRAM," to the TSs. All existing references to the "lnservice Testing Program" in the TS SRs, for each facility, are replaced with "INSERVICE TESTING PROGRAM" so that the SRs refer to the new definition in lieu of the deleted program.

The licensee's application, as supplemented, also requested similar amendments for Nine Mile Point Unit No. 1, and Three Mile Island Nuclear Station, Unit 1. However, the U.S. Nuclear Regulatory Commission (NRC or Commission) staff's safety evaluations for these plants are provided separately, and they are not considered in this safety evaluation.

2.0 REGULATORY EVALUATION

2.1 Description of lnservice Testing Requirements and TSTF-545 An inservice test is a test to assess the operational readiness of a structure, system, or component after first electrical generation by nuclear heat. The American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) provides requirements for inservice testing of certain components in light-water nuclear power plants. The ASME OM Code identifies the components subject to the testing (i.e., pumps, valves, pressure relief devices, and dynamic restraints), responsibilities, methods, intervals, para.meters to be measured and evaluated, criteria for evaluating results, corrective actions, personnel qualification, and recordkeeping. Title 10 of the Code of Federal Regulations (10 CFR), paragraph 50.55a(f), "lnservice testing requirements," requires that inservice testing of certain ASME Code Class 1, 2, and 3 components must meet the requirements of the ASME OM Code and applicable addenda. The facility's TSs also prescribe inservice testing requirements and frequencies for ASME Code Class 1, 2, and 3 components.

The regulation in 10 CFR 50.55a(f)(5)(ii) states, in part: "If a revised inservice test program for a facility conflicts with the technical specifications for the facility, the licensee must apply to the Commission for amendment of the technical specifications to conform the technical specifications to the revised program." TSTF-545, Revision 3, provides guidance to licensees on how to request license amendments that would eliminate conflicting requirements between 10 CFR 50.55a, "Codes and standards," and the TSs. TSTF-545, Revision 3, proposes elimination of the lnservice Testing Program from the Administrative Controls section of the TSs. The TSs contain surveillances that require testing or test intervals in accordance with the lnservice Testing Program. The elimination of the lnservice Testing Program from the TSs could cause uncertainty regarding the correct application of these SRs. Therefore, TSTF 545, Revision 3, also proposes adding a new definition, "INSERVICE TESTING PROGRAM," to the TSs, which would be defined as "the licensee program that fulfills the requirements of 10 CFR 50.55a(f)." TSTF-545, Revision 3, proposes replacement of existing uses of the term, "lnservice Testing Program," with the defined term, as denoted by capitalized letters, throughout the TSs.

The NRC approved TSTF-545, Revision 3, by letter dated December 11, 2015 (ADAMS Package Accession No. ML15317A071), and published a notice of availability in the Federal Register(FR) on March 28, 2016 (81FR17208).

2.2 Technical Specifications Changes The licensee proposed to delete the lnservice Testing Program from TS Section 5.5 for each facility and replace it with the word "DELETED." Currently, TS 5.5.8, "lnservice Testing Program," for Braidwood, Units 1 and 2, states:

This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following:

a. Testing frequencies applicable to the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as follows:

ASME OM Code and applicable Addenda terminology for Required Frequencies inservice testing for performing inservice activities testing activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days;

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the lnservice Testing Program for performing inservice testing activities;

c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and

d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any Technical Specification.

TS 5.5.8.b, which refers to SR 3.0.2, allows an extension of inservice testing intervals by up to 25 percent. If it is discovered that a surveillance associated with an inservice testing activity was not performed within the required interval, SR 3.0.3 allows the licensee to delay declaring the associated limiting condition for operation not met in order to perform the missed surveillance. The licensee did not request changes to SR 3.0.2 or SR 3.0.3.

The current lnservice Testing Program requirements in TS 5.5.8 for Byron, Units 1 and 2; TS 5.5.6 for Dresden, Units 2 and 3; TS 5.5.8 for Calvert Cliffs, Units 1 and 2; TS 5.5.6 for Clinton; TS 5.5.7 for LaSalle, Units 1 and 2; TS 5.5.6 for Nine Mile Point, Unit 2; TS 5.5.6 for Peach Bottom, Units 2 and 3; TS 5.5.7 for Ginna; and TS 5.5.6 for Quad Cities, Units 1 and 2, are similar to TS 5.5.8 for Braidwood, Units 1 and 2. 1 Aside from the TS numbering, the primary differences from the Braidwood TS 5.5.8 are the following:

1. Format and grammar (e.g., use of 'TS" versus "Technical Specification").

2. The Dresden, LaSalle, Nine Mile Point Unit 2, and Quad Cities TSs refer to "pumps and "valves" instead of "components" in the first sentence.

3. The Ginna TS states "components including applicable supports" instead of just "components" in the first sentence.

4. The Dresden, LaSalle, Nine Mile Point Unit 2, and Quad Cities TSs do not have the second sentence, which states: "The program shall include the following:".

5. The Dresden, LaSalle, and Quad Cities TSs for the lnservice Testing Program also require that when the ASME OM Code and applicable Addenda specify "Every 1 This safety evaluation uses the phrase 'TS 5.5.8 (or equivalent)" to refer to these current TSs for the lnservice Testing Program.

48 months" the required frequency for performing the testing activity is "At least once per 1461 days."

For each facility, the licensee requested to revise the Definitions section of TSs by adding the term, "INSERVICE TESTING PROGRAM," with the following definition: "The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f)." The licensee also requested that all existing occurrences of "lnservice Testing Program" in TS SRs, for each facility, be replaced with "INSERVICE TESTING PROGRAM," so that the SRs refer to the new definition in lieu of the deleted program.

2.3 Regulatory Requirements and Guidance The NRC staff considered the following regulatory requirements, guidance, and licensing information during its review of the proposed changes:

Technical Specifications Paragraph 50.36(c) of 10 CFR requires TSs to include the following categories: (1) safety limits, limiting safety systems settings, and control settings; (2) limiting conditions for operation; (3) SRs; (4) design features; (5) administrative controls; (6) decommissioning; (7) initial notification; and (8) written reports. Section 50.36(c)(3) of 10 CFR states that "[s]urveillance requirements 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." Section 50.36(c)(5) of 10 CFR states that "[a]dministrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner."

The NRC staff's guidance for review of the TSs is in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor]

Edition," Chapter 16, "Technical Specifications," Revision 3, dated March 2010 (ADAMS Accession No. ML100351425). As described therein, as part of the regulatory standardization effort, the staff has prepared improved STSs for each of the LWR nuclear steam supply systems and associated balance-of-plant equipment systems. The licensee's proposed amendment is based on TSTF-545, Revision 3, which is an NRG-approved change to the improved STSs.

The staff's review includes consideration of whether the proposed changes are consistent with TSTF-545, Revision 3. Special attention is given to TS provisions that depart from the improved STSs, as modified by NRG-approved TSTF travelers, to determine whether proposed differences are justified by uniqueness in plant design or other considerations so that 10 CFR 50.36 is met. In addition, the guidance states that comparing the change to previous STS can help clarify the TS intent.

lnseNice Testing Pursuant to 10 CFR 50.54, "Conditions of licenses," the applicable requirements of 10 CFR 50.55a are conditions of every nuclear power reactor operating license issued under 10 CFR Part 50. These requirements include inservice testing of pumps and valves at nuclear power reactors in accordance with the ASME OM Code as specified in 10 CFR 50.55a(f). The regulations in 10 CFR 50.55a(f) state, in part:

Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME BPV [Boiler and Pressure Vessel] Code and ASME Code for Operation and Maintenance of Nuclear Power Plants as specified in this paragraph. Each operating license for a boiling or pressurized water-cooled nuclear facility is subject to the following conditions

[referring to 10 CFR 50.55a(f)(1) through (f)(6)] ....

The ASME OM Code is a consensus standard, which is incorporated by reference into 10 CFR 50.55a. During the incorporation process, the NRC staff reviewed the ASME OM Code requirements for technical sufficiency and found that the ASME OM Code inservice testing program requirements were suitable for incorporation into the NRC's rules.

The regulation in 10 CFR 50.55(a)(f)(5)(ii) states, in part: "If a revised inservice test program for a facility conflicts with the technical specifications for the facility, the licensee must apply to the Commission for amendment of the technical specifications to conform the technical specifications to the revised program."

NUREG-1482, Revision 2, "Guidelines for lnservice Testing at Nuclear Power Plants: lnservice Testing of Pumps and Valves and lnservice Examination and Testing of Dynamic Restraints (Snubbers) at Nuclear Power Plants," Final Report, October 2013 (ADAMS Accession No. ML13295A020), provides guidance for the inservice testing of pumps and valves.

NUREG-0800, Section 3.9.6, "Functional Design, Qualification, and lnservice Testing Programs for Pumps, Valves, and Dynamic Restraints," Revision 3, March 2007 (ADAMS Accession No. ML070720041 ), provides guidance and acceptance criteria for the NRC staff review of the inservice testing program for pumps and valves.

3.0 TECHNICAL EVALUATION

The NRC staff evaluated the licensee's application, as supplemented, to determine if the proposed changes are consistent with the guidance, regulations, and licensing information discussed in Section 2.3 of this safety evaluation. In determining whether an amendment to a license will be issued, the Commission is guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. Among the considerations are whether the TSs, as amended, would provide the necessary administrative controls per 10 CFR 50.36(c)(5) (i.e., provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner). In making its determination as to whether to amend the license, the staff considered those regulatory requirements that are automatically conditions of the license through 10 CFR 50.54. Where the regulations already condition the license, there is no need for a duplicative requirement in the TSs; the regulations provide the necessary reasonable assurance of the health and safety of the public.

3.1 Deletion of the lnservice Testing Program from the TSs TS 5.5.8 (or equivalent) requires the licensee to have an inservice testing program that provides controls for inservice testing of ASME Code Class 1, 2, and 3 components (i.e., pumps and valves). Through 10 CFR 50.54, the applicable requirements of 10 CFR 50.55a are conditions of every nuclear power reactor operating license issued under 10 CFR Part 50. These requirements include 10 CFR 50.55a(f), which specifies the requirements for the inservice testing of pumps and valves. Therefore, requiring the licensee to have an inservice testing

program in TSs is duplicative of the license condition in 10 CFR 50.54. Thus, with the proposed TS changes, the licensee will still be required to maintain an inservice testing program in accordance with the ASME OM Code, as specified in 10 CFR 50.55a(f). For the reasons explained below, it is not necessary to have additional administrative controls in the TSs relating to the inservice testing program to assure operation of the facility in a safe manner.

Consideration of TS 5.5.8.a (or equivalent)

The ASME OM Code requires testing to normally be performed within certain time periods.

TS 5.5.8.a (or equivalent) sets inservice testing frequencies more precisely than those specified in the ASME OM Code and applicable addenda (e.g., "at least once per 31 days" contrasted with "monthly"). However, the NRC staff determined that the more precise inservice testing frequencies are not necessary to assure operation of the facility in a safe manner. Therefore, the staff determined that deletion of TS 5.5.8.a (or equivalent) is acceptable.

Consideration of TS 5. 5. 8. b (or equivalent)

TS 5.5.8.b (or equivalent) allows the licensee to extend, by up to 25 percent, the interval between inservice testing activities, as required by TS 5.5.8.a (or equivalent) and for other normal and accelerated frequencies specified as 2 years or less in the inservice testing program. For each facility, the NRC has authorized the use of ASME Code Case OMN-20, "lnservice Test Frequency," or similar alternatives to the ASME OM Code. 2 Like TS 5.5.8.b (or equivalent), these NRG-authorized alternatives also permit the licensee to extend the inservice testing intervals specified in the ASME OM Code by up to 25 percent.

The NRC staff determined that the TS 5.5.8.b (or equivalent) allowance to extend inservice testing intervals is not needed to assure operation of the facility in a safe manner. Therefore, the NRC staff determined that deletion of TS 5.5.8.b (or equivalent) is acceptable. The deletion of TS 5.5.8.b (or equivalent) does not impact the licensee's ability to extend inservice testing intervals using Code Case OMN-20 or similar alternatives, as authorized by the NRC.

Consideration of TS 5. 5. 8. c (or equivalent)

TS 5.5.8.c (or equivalent) allows the licensee to use SR 3.0.3 when it discovers that an SR associated with an inservice test was not performed within its specified frequency. SR 3.0.3 allows the licensee to delay declaring a limiting condition for operation not met in order to perform the missed surveillance. The use of SR 3.0.3 for inservice tests is limited to those inservice tests required by an SR In accordance with 10 CFR 50.55a, the licensee may also request relief from the ASME OM Code requirements to address issues associated with a missed inservice test. Deletion of TS 5.5.8.c (or equivalent) does not change any of these requirements, and SR 3.0.3 will continue to apply to those inservice tests required by SRs.

Based on the above, the NRC staff determined that deletion of TS 5.5.8.c (or equivalent) is acceptable.

2 The NRC authorizations were made by letters dated February 14 and October 31, 2013; September 24, 2014; February 26, 2016; and February 21, 2017 (ADAMS Accession Nos. ML13042A348, ML13297A515, ML14247A555, ML16022A135, and ML17046A286, respectively).

Consideration of TS 5. 5. 8. d (or equivalent)

TS 5.5.8.d (or equivalent) states that nothing in the ASME OM Code shall be construed to supersede the requirements of any TS. However, the regulations in 10 CFR 50. 55a(f)(5)(ii) address what to do if a revised inservice testing program for a facility conflicts with the TSs for the facility; they require the licensee to apply for an amendment to the TSs to conform the TSs to the revised program at least 6 months prior to the start of the period for which the provisions become applicable. Accordingly, there is no need for a TS stating how to address conflicts between the TSs and the inservice testing program because the regulations specify how conflicts must be resolved.

Conclusion Regarding Deletion of TS 5.5.8 (or equivalent)

The NRC staff determined that the requirements currently in TS 5.5.8 (or equivalent) are not necessary to assure operation of the facility in a safe manner. Based on this evaluation, the staff concludes that deletion of TS 5.5.8 (or equivalent) from the licensee's TSs for each facility is acceptable, because TS 5.5.8 (or equivalent) is not required by 10 CFR 50.36(c)(5).

3.2 Definition of INSERVICE TESTING PROGRAM and Revision to SRs The licensee proposes to revise the TS Definitions section to include the term, "INSERVICE TESTING PROGRAM," with the following definition: "The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f)." The proposed definition of the INSERVICE TESTING PROGRAM is consistent with the definition in TSTF-545, Revision 3. The definition is acceptable to the NRC staff because it correctly refers to the inservice testing requirements in 10 CFR 50.55a(f).

The licensee requested that all existing references to the "lnservice Testing Program" in SRs be revised to "INSERVICE TESTING PROGRAM" to reference the new TS defined term in lieu of the deleted program. The proposed change is consistent with the intent of TSTF-545, Revision 3, to replace the current references in SRs with the new definition. The NRC staff verified that for each SR reference to the "lnservice Testing Program," the licensee proposed to change the reference to "INSERVICE TESTING PROGRAM." The proposed change does not alter how the SR testing is performed. However, the inservice testing frequencies could change because the TSs will no longer include the more precise test frequencies in TS 5.5.8.a (or equivalent). As discussed in Section 3.1 of this safety evaluation, the staff determined that the TSs do not need to include the more precise testing frequencies currently in TS 5.5.8.a (or equivalent). Based on its review, the staff determined that revising the SRs to refer to the new definition is acceptable because these SRs will continue to be performed in accordance with the requirements of 10 CFR 50.55a(f). The staff also determined that, with the proposed changes that allow less-precise testing frequencies, 10 CFR 50.36(c)(3) will continue to be met because the SRs will continue to ensure 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.

3.3 Deviations from TSTF-545 The licensee's October 6, 2016, letter identified the following deviations from the TSTF-545, Revision 3:

1. TSTF-545, Revision 3, completely deletes TS 5.5.8 (or equivalent) from the TSs and renumbers the subsequent TS programs. The licensee proposes to delete the content of TS 5.5.8 (or equivalent), but retains the TS number, and adds the word "DELETED."

The licensee did not propose to renumber the subsequent TS programs.

2. For each facility, some of the numbering for SRs that are modified does not match the numbering in TSTF-545, Revision 3. However, the licensee stated that the SRs are equivalent.

The NRC staff finds that the proposed deviations are editorial in nature and the licensee's proposed TS changes remain consistent with the intent of TSTF-545, Revision 3. Therefore, the staff finds that the licensee's proposed TS changes are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Illinois, Maryland, Pennsylvania, and New York State officials were notified of the proposed issuance of the amendments on March 16, 2017. The State officials had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to the installation or use of facility components located within the restricted areas as defined in 10 CFR Part 20 and changes SRs.

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, which was published in the FR on November 8, 2016 (81 FR 78648), that the amendments involve no significant hazards consideration, and there has been no public comment on such finding. 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 amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: Blake Purnell, NRR Date of issuance: May 26, 2017

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 227 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-63 EXELON GENERATION COMPANY, LLC NINE MILE POINT NUCLEAR STATION UNIT NO. 1 DOCKET NO. 50-220

1.0 INTRODUCTION

By application dated July 26, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16209A218), as supplemented by letter dated October 6, 2016 (ADAMS Accession No. ML16280A402), Exelon Generation Company, LLC (the licensee) requested changes to the technical specifications (TSs) for Nine Mile Point Nuclear Station Unit No. 1 (NMP-1 ). Specifically, the licensee requested changes to the TSs consistent with Technical Specifications Task Force (TSTF) Standard Technical Specifications (STS) Change Traveler TSTF-545, Revision 3, "TS lnservice Testing Program Removal & Clarify SR

[Surveillance Requirement] Usage Rule Application to Section 5.5 Testing," dated October 21, 2015 (ADAMS Accession No. ML15294A555). The licensee's October 6, 2016, letter revised the scope of the TS changes requested and withdrew the proposed TS changes originally requested in the July 26, 2016, application.

The licensee's proposed changes to delete NMP-1 TS 6.5.4, "lnservice Testing Program," and adds a new defined term, "INSERVICE TESTING PROGRAM," to the TSs. The reference to "Specification 6.5.4" in SR 4.2.7 is replaced with "the INSERVICE TESTING PROGRAM" so that SR 4.2. 7 refer to the new definition in lieu of the deleted program.

The licensee's application, as supplemented, also requested similar amendments for other Exelon Generation Company, LLC facilities. 1 However, the U.S. Nuclear Regulatory Commission (NRC or Commission) staff's safety evaluations for these other facilities are provided separately, and they are not considered in this safety evaluation.

1 The other facilities are Braidwood Station, Units 1 and 2; Byron Station, Unit Nos. 1 and 2; Calvert Cliffs Nuclear Power Plant, Units 1 and 2; Clinton Power Station, Unit No. 1; Dresden Nuclear Power Station, Units 2 and 3; LaSalle County Station, Units 1 and 2; Nine Mile Point Nuclear Station, Unit 2; Peach Bottom Atomic Power Station, Units 2 and 3; Quad Cities Nuclear Power Station, Units 1 and 2; R. E.

Ginna Nuclear Power Plant; and Three Mile Island Nuclear Station, Unit 1.

Enclosure 21

2.0 REGULATORY EVALUATION

2.1 Description of lnservice Testing Requirements and TSTF-545 An inservice test is a test to assess the operational readiness of a structure, system, or component after first electrical generation by nuclear heat. The American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) provides requirements for inservice testing of certain components in light-water nuclear power plants. The ASME OM Code identifies the components subject to the testing (i.e., pumps, valves, pressure relief devices, and dynamic restraints), responsibilities, methods, intervals, parameters to be measured and evaluated, criteria for evaluating results, corrective actions, personnel qualification, and recordkeeping. Title 10 of the Code of Federal Regulations (10 CFR), paragraph 50.55a(f), "lnservice testing requirements," requires that inservice testing of certain ASME Code Class 1, 2, and 3 components must meet the requirements of the ASME OM Code and applicable addenda. The facility's TSs also prescribe inservice testing requirements and frequencies for ASME Code Class 1, 2, and 3 components.

The regulation in 10 CFR 50.55a(f)(5)(ii) states, in part: "If a revised inservice test program for a facility conflicts with the technical specifications for the facility, the licensee must apply to the Commission for amendment of the technical specifications to conform the technical specifications to the revised program." TSTF-545, Revision 3, provides guidance to licensees on how to request license amendments that would eliminate conflicting requirements between 10 CFR 50.55a, "Codes and standards," and the TSs. TSTF-545, Revision 3, proposes elimination of the lnservice Testing Program from the Administrative Controls section of the TSs. The TSs contain surveillances that require testing or test intervals in accordance with the lnservice Testing Program. The elimination of the lnservice Testing Program from the TSs could cause uncertainty regarding the correct application of these SRs. Therefore, TSTF 545, Revision 3, also proposes adding a new definition, "INSERVICE TESTING PROGRAM," to the TSs, which would be defined as "the licensee program that fulfills the requirements of 10 CFR 50.55a(f)." TSTF-545, Revision 3, proposes replacement of existing uses of the term, "lnservice Testing Program," with the defined term, as denoted by capitalized letters, throughout the TSs.

The NRC approved TSTF-545, Revision 3, by letter dated December 11, 2015 (ADAMS Package Accession No. ML15317A071 ), and published a notice of availability in the Federal Register(FR) on March 28, 2016 (81FR17208).

2.2 Proposed Technical Specifications Changes The licensee requested to delete TS 6.5.4 from the Administrative Controls section of TSs and replace it with the word "DELETED." TS 6.5.4 currently states:

This program provides controls for inservice testing of Quality Group A, B, and C pumps and valves.

a. lnservice testing of Quality Group A, B, and C pumps and valves shall be performed in accordance with requirements for American Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 components specified in the applicable Edition and Addenda of the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code), subject to the applicable provisions of 10CFR50.55a;

b. The provisions of Specification 4.0.2 are applicable to the normal and accelerated testing frequencies specified as 2 years or less in the lnservice Testing Program for performing inservice testing activities;

c. The provisions of Specification 4.0.3 are applicable to inservice testing activities; and

d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any Technical Specification.

TS 6.5.4.b, which refers to Specification 4.0.2 (SR 4.0.2), allows an extension of inservice testing intervals by up to 25 percent. If it is discovered that a surveillance associated with an inservice testing activity was not performed within the required interval, Specification 4.0.3 (SR 4.0.3) allows the licensee to delay declaring the associated limiting condition for operation

• not met in order to perform the missed surveillance. The licensee did not request changes to SR 4.0.2 or SR 4.0.3.

The licensee requested to revise the Definitions section of TSs by adding Definition 1.33, "INSERVICE TESTING PROGRAM," with the following definition: "The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f)." The licensee also requested that the reference to "Specification 6.5.4" in SR 4.2.7 be replaced with "INSERVICE TESTING PROGRAM," so that SR 4.2.7 refer to the new definition in lieu of the deleted program.

2.3 Regulatory Requirements and Guidance The NRC staff considered the following regulatory requirements, guidance, and licensing information during its review of the proposed changes:

Technical Specifications Paragraph 50.36(c) of 10 CFR requires TSs to include the following categories: (1) safety limits, limiting safety systems settings, and control settings; (2) limiting conditions for operation; (3) SRs; (4) design features; (5) administrative controls; (6) decommissioning; (7) initial notification; and (8) written reports. Section 50.36(c)(3) of 10 CFR states that "[s]urveillance requirements 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." Section 50.36(c)(5) of 10 CFR states that "[a]dministrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner."

The NRC staff's guidance for review of the TSs is in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor]

Edition," Chapter 16, "Technical Specifications," Revision 3, dated March 2010 (ADAMS Accession No. ML100351425). As described therein, as part of the regulatory standardization effort, the staff has prepared improved STSs for each of the LWR nuclear steam supply systems and associated balance-of-plant equipment systems. The licensee's proposed amendment is based on TSTF-545, Revision 3, which is an NRG-approved change to the improved STSs.

The staff's review includes consideration of whether the proposed changes are consistent with TSTF-545, Revision 3. Special attention is given to TS provisions that depart from the improved

STSs, as modified by NRG-approved TSTF travelers, to determine whether proposed differences are justified by uniqueness in plant design or other considerations so that 10 CFR 50.36 is met. In addition, the guidance states that comparing the change to previous STS can help clarify the TS intent.

lnservice Testing Pursuant to 10 CFR 50.54, "Conditions of licenses," the applicable requirements of 10 CFR 50.55a are conditions of every nuclear power reactor operating license issued under 10 CFR Part 50. These requirements include inservice testing of pumps and valves at nuclear power reactors in accordance with the ASME OM Code as specified in 10 CFR 50.55a(f). The regulations in 10 CFR 50.55a(f) state, in part:

Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME BPV [Boiler and Pressure Vessel] Code and ASME Code for Operation and Maintenance of Nuclear Power Plants as specified in this paragraph. Each operating license for a boiling or pressurized water-cooled nuclear facility is subject to the following conditions

[referring to 10 CFR 50.55a(f)(1) through (f)(6)] ....

The ASME OM Code is a consensus standard, which is incorporated by reference into 10 CFR 50.55a. During the incorporation process, the NRC staff reviewed the ASME OM Code requirements for technical sufficiency and found that the ASME OM Code inservice testing program requirements were suitable for incorporation into the NRC's rules.

The regulation in 10 CFR 50.55(a)(f)(5)(ii) states, in part: "If a revised inservice test program for a facility conflicts with the technical specifications for the facility, the licensee must apply to the Commission for amendment of the technical specifications to conform the technical specifications to the revised program."

NUREG-1482, Revision 2, "Guidelines for lnservice Testing at Nuclear Power Plants: lnservice Testing of Pumps and Valves and lnservice Examination and Testing of Dynamic Restraints (Snubbers) at Nuclear Power Plants," Final Report, October 2013 (ADAMS Accession No. ML13295A020), provides guidance for the inservice testing of pumps and valves.

NUREG-0800, Section 3.9.6, "Functional Design, Qualification, and lnservice Testing Programs for Pumps, Valves, and Dynamic Restraints," Revision 3, March 2007 (ADAMS Accession No. ML070720041 ), provides guidance and acceptance criteria for the NRC staff review of the inservice testing program for pumps and valves.

3.0 TECHNICAL EVALUATION

The NRC staff evaluated the licensee's application, as supplemented, to determine if the proposed changes are consistent with the guidance, regulations, and licensing information discussed in Section 2.3 of this safety evaluation. In determining whether an amendment to a license will be issued, the Commission is guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. Among the considerations are whether the TSs, as amended, would provide the necessary administrative controls per 10 CFR 50.36(c)(5) (i.e., provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner). In making its determination as to whether to amend the license, the staff

considered those regulatory requirements that are automatically conditions of the license through 10 CFR 50.54. Where the regulations already condition the license, there is no need for a duplicative requirement in the TSs; the regulations provide the necessary reasonable assurance of the health and safety of the public.

3.1 Deletion of the lnservice Testing Program from the TSs TS 6.5.4 requires the licensee to have an inservice testing program that provides controls for inservice testing of Quality Group A, B, and C pumps and valves. TS 6.5.4.a requires that these pumps and valves be tested in accordance with requirements for ASME Code Class 1, 2, and 3 components specified in the applicable Edition and Addenda of the ASME OM Code, in accordance with 10 CFR 50.55a. Through 10 CFR 50.54, the applicable requirements of 10 CFR 50.55a are conditions of every nuclear power reactor operating license issued under 10 CFR Part 50. These requirements include 10 CFR 50.55a(f), which specifies the requirements for the inservice testing of pumps and valves. Therefore, requiring the licensee to have an inservice testing program in TSs is duplicative of the license condition in 10 CFR 50.54.

Thus, with the proposed TS changes, the licensee will still be required to maintain an inservice testing program in accordance with the ASME OM Code, as specified in 10 CFR 50.55a(f).

Based on this, the NRC staff determined that deletion of TS 6.5.4.a is acceptable. For the reasons explained below, it is not necessary to have additional administrative controls in the TSs relating to the inservice testing program to assure operation of the facility in a safe manner.

Consideration of TS 6.5.4.b TS 6.5.4.b allows the licensee to extend, by up to 25 percent, the interval between inservice testing activities and for other normal and accelerated frequencies specified as 2 years or less in the inservice testing program. Similar to TS 6.5.4.b, the NRC authorization of ASME Code Case OMN-20, "lnservice Test Frequency," by letter dated February 21, 2017 (ADAMS Accession No. ML17046A286), also permits the licensee to extend the inservice testing intervals specified in the ASME OM Code by up to 25 percent.

The NRC staff determined that the TS 6.5.4.b allowance to extend inservice testing intervals is not needed to assure operation of the facility in a safe manner. Therefore, the staff determined that deletion of TS 6.5.4.b is acceptable. The deletion of TS 6.5.4.b does not impact the licensee's ability to extend inservice testing intervals using Code Case OMN-20, as authorized by the NRC.

Consideration of TS 6.5.4.c TS 6.5.4.c allows the licensee to use SR 4.0.3 when it discovers that an SR associated with an inservice test was not performed within its specified frequency. SR 4.0.3 allows the licensee to delay declaring a limiting condition for operation not met in order to perform the missed surveillance. The use of SR 4.0.3 for inservice tests is limited to those inservice tests required by an SR. In accordance with 10 CFR 50.55a, the licensee may also request relief from the ASME OM Code requirements to address issues associated with a missed inservice test.

Deletion of TS 6.5.4.c does not change any of these requirements, and SR 4.0.3 will continue to apply to those inservice tests required by SRs. Based on the above, the NRC staff determined that deletion of TS 6.5.4.c is acceptable.

Consideration of TS 6.5.4.d TS 6.5.4.d states that nothing in the ASME OM Code shall be construed to supersede the requirements of any TS. However, the regulations in 10 CFR 50.55a(f)(5)(ii) address what to do if a revised inservice testing program for a facility conflicts with the TSs for the facility; they require the licensee to apply for an amendment to the TSs to conform the TSs to the revised program at least 6 months prior to the start of the period for which the provisions become applicable. Accordingly, there is no need for a TS stating how to address conflicts between the TSs and the inservice testing program because the regulations specify how conflicts must be resolved.

Conclusion Regarding Deletion of TS 6. 5. 4 The NRC staff determined that the requirements currently in TS 6.5.4 are not necessary to assure operation of the facility in a safe manner. Based on this evaluation, the staff concludes that deletion of TS 6.5.4 from the licensee's TSs is acceptable, because TS 6.5.4 is not required by 10 CFR 50.36(c)(5).

3.2 Definition of INSERVICE TESTING PROGRAM and Revision to SR 4.2.7 The licensee proposes to revise the TS Definitions section to add Definition 1.33, "INSERVICE TESTING PROGRAM," with the following definition: "The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f)." The proposed definition of the INSERVICE TESTING PROGRAM is consistent with the definition in TSTF-545, Revision 3. The definition is acceptable to the NRC staff because it correctly refers to the inservice testing requirements in 10 CFR 50.55a(f).

The licensee requested that the reference to "Specification 6.5.4" in SR 4.2.7 be replaced with "INSERVICE TESTING PROGRAM," so that SR 4.2.7 refers to the new definition in lieu of the deleted program. The proposed change is consistent with the intent of TSTF-545, Revision 3, to replace the current references in SRs with the new definition. The NRC staff verified that the TSs do not contain any other references to "Specification 6.5.4" or the "lnservice Testing Program." The proposed change does not alter how the SR testing is performed. Based on its review, the staff determined that revising SR 4.2. 7 to refer to the new definition is acceptable because SR 4.2. 7 will continue to be performed in accordance with the requirements of 10 CFR 50.55a(f). The staff also determined that, with the proposed changes, 10 CFR 50.36(c)(3) will continue to be met because the SRs will continue to ensure 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.

3.3 Deviations from TSTF-545 In its application, the licensee identified the following deviations from TSTF-545, Revision 3:

1. TSTF-545, Revision 3, completely deletes the lnservice Test Program from the Administrative Controls section of the TSs and renumbers the subsequent TS programs.

The licensee proposes to delete the content of TS 6.5.4, but retains the TS number, and adds the word "DELETED." The licensee did not propose to renumber the subsequent TS programs.

2. NMP-1 has custom TSs that differ in numbering, titles, and SRs from the improved STS on which TSTF-545, Revision 3, was based. Administrative Controls are in NMP-1 TS Section 6.0, instead of Section 5.0. NMP-1 SR 4.2.7 refers to "Specification 6.5.4,"

instead of the "lnservice Testing Program." In addition, many of the SRs identified in TSTF-545, Revision 3, are not applicable to NMP-1.

The NRC staff finds that the proposed deviations are editorial in nature and the licensee's proposed TS changes remain consistent with the intent of TSTF-545, Revision 3. Therefore, the staff finds that the licensee's proposed TS changes are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New York State official was notified of the proposed issuance of the amendment on March 16, 2017. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes SRs.

The NRC staff has determined that the amendment involves 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, which was published in the FR on November 8, 2016 (81 FR 78648), that the amendment involves no significant hazards consideration, and there has been no public comment on such finding. Accordingly, the amendment meets 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 amendment.

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 amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: Blake Purnell, NRR Date of issuance: May 2 6, 2O1 7

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 290 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-50 EXELON GENERATION COMPANY. LLC THREE MILE ISLAND NUCLEAR STATION, UNIT 1 DOCKET NO. 50-289

1.0 INTRODUCTION

By application dated July 26, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16209A218), as supplemented by letter dated October 6, 2016 (ADAMS Accession No. ML16280A402), Exelon Generation Company, LLC (the licensee) requested changes to the technical specifications (TSs) for Three Mile Island Nuclear Station, Unit 1 (TMl-1 ). Specifically, the licensee requested changes to the TSs consistent with Technical Specifications Task Force (TSTF) Standard Technical Specifications (STS) Change Traveler TSTF-545, Revision 3, TS lnservice Testing Program Removal & Clarify SR

[Surveillance Requirement] Usage Rule Application to Section 5.5 Testing, dated October 21, 2015 (ADAMS Accession No. ML15294A555). The licensee's October 6, 2016, letter revised the scope of the TS changes requested and withdrew the proposed TS changes originally requested in the July 26, 2016, application.

The licensee's proposed changes delete the lnservice Testing Program requirements from TMl-1 TS 4.2, "Reactor Coolant System lnservice and Testing, and adds a new defined term, "INSERVICE TESTING PROGRAM," to the TSs. Remaining references to the "lnservice Testing Program" in the TMl-1 TSs are replaced with "INSERVICE TESTING PROGRAM" so that TSs refer to the new definition in lieu of the deleted program.

The licensee's application, as supplemented, also requested similar amendments for other Exelon Generation Company, LLC facilities. 1 However, the U.S. Nuclear Regulatory Commission (NRC or Commission) staff's safety evaluations for these other facilities are provided separately, and they are not considered in this safety evaluation.

• 1 The other facilities are Braidwood Station, Units 1 and 2; Byron Station, Unit Nos. 1 and 2; Calvert Cliffs Nuclear Power Plant, Units 1 and 2; Clinton Power Station, Unit No. 1; Dresden Nuclear Power Station, Units 2 and 3; LaSalle County Station, Units 1 and 2; Nine Mile Point Nuclear Station, Units 1 and 2; Peach Bottom Atomic Power Station, Units 2 and 3; Quad Cities Nuclear Power Station, Units 1 and 2; and R. E. Ginna Nuclear Power Plant.

Enclosure 22

2.0 REGULATORY EVALUATION

2.1 Description of lnservice Testing Requirements and TSTF-545 An inservice test is a test to assess the operational readiness of a structure, system, or component after first electrical generation by nuclear heat. The American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) provides requirements for inservice testing of certain components in light-water nuclear power plants. The ASME OM Code identifies the components subject to the testing (i.e., pumps, valves, pressure relief devices, and dynamic restraints), responsibilities, methods, intervals, parameters to be measured and evaluated, criteria for evaluating results, corrective actions, personnel qualification, and recordkeeping. Title 1O of the Code of Federal Regulations (10 CFR), paragraph 50.55a(f), "lnservice testing requirements," requires that inservice testing of certain ASME Code Class 1, 2, and 3 components must meet the requirements of the ASME OM Code and applicable addenda. The facility's TSs also prescribe inservice testing requirements and frequencies for ASME Code Class 1, 2, and 3 components.

The regulation in 10 CFR 50.55a(f)(5)(ii) states, in part: "If a revised inservice test program for a facility conflicts with the technical specifications for the facility, the licensee must apply to the Commission for amendment of the technical specifications to conform the technical specifications to the revised program." TSTF-545, Revision 3, provides guidance to licensees on how to request license amendments that would eliminate conflicting requirements between 10 CFR 50.55a, "Codes and standards," and the TSs. TSTF-545, Revision 3, proposes elimination of the lnservice Testing Program from the Administrative Controls section of the TSs. The TSs contain surveillances that require testing or test intervals in accordance with the lnservice Testing Program. The elimination of the lnservice Testing Program from the TSs could cause uncertainty regarding the correct application of these SRs. Therefore, TSTF 545, Revision 3, also proposes adding a new definition, "INSERVICE TESTING PROGRAM," to the TSs, which would be defined as "the licensee program that fulfills the requirements of 10 CFR 50.55a(f)." TSTF-545, Revision 3, proposes replacement of existing uses of the term, "lnservice Testing Program," with the defined term, as denoted by capitalized letters, throughout the TSs.

The NRC approved TSTF-545, Revision 3, by letter dated December 11, 2015 (ADAMS Package Accession No. ML15317A071), and published a notice of availability in the Federal Register (FR) on March 28, 2016 (81 FR 17208).

2.2 Proposed Technical Specifications Changes The licensee requested to delete the lnservice Testing Program requirements from TS 4.2, by deleting TS 4.2.2 and replacing it with the word "DELETED." In addition, references to inservice testing (IST) are removed from the Applicability and Objective sections of TS 4.2. TS 4.2.2 currently states:

IST of ASME Code Class 1, Class 2 and Class 3 pumps and valves shall be performed in accordance with the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as required by 10 CFR 50, Section 50.55a, except where specific written relief has been granted by the NRC.

The licensee requested to revise the Definitions section of TSs by adding Definition 1.27, "INSERVICE TESTING PROGRAM, with the following definition: 'The INSERVICE TESTING

PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f)." The licensee also requested that all existing occurrences of "lnservice Testing Program" in SRs (TMl-1 TS Section 4, "Surveillance Standards") be replaced with "INSERVICE TESTING PROGRAM," so that the SRs refer to the new definition in lieu of the deleted program.

2.3 Regulatory Requirements and Guidance The NRC staff considered the following regulatory requirements, guidance, and licensing information during its review of the proposed changes:

Technical Specifications Paragraph 50.36(c) of 10 CFR requires TSs to include the following categories: (1) safety limits, limiting safety systems settings, and control settings; (2) limiting conditions for operation; (3) SRs; (4) design features; (5) administrative controls; (6) decommissioning; (7) initial notification; and (8) written reports. Section 50.36(c)(3) of 10 CFR states that "[s]urveillance requirements 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." Section 50.36(c)(5) of 10 CFR states that "[a]dministrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner."

The NRC staff's guidance for review of the TSs is in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor]

Edition," Chapter 16, "Technical Specifications," Revision 3, dated March 2010 (ADAMS Accession No. ML100351425). As described therein, as part of the regulatory standardization effort, the staff has prepared improved STSs for each of the LWR nuclear steam supply systems and associated balance-of-plant equipment systems. The licensee's proposed amendment is based on TSTF-545, Revision 3, which is an NRG-approved change to the improved STSs.

The staff's review includes consideration of whether the proposed changes are consistent with TSTF-545, Revision 3. Special attention is given to TS provisions that depart from the improved STSs, as modified by NRG-approved TSTF travelers, to determine whether proposed differences are justified by uniqueness in plant design or other considerations so that 10 CFR 50.36 is met. In addition, the guidance states that comparing the change to previous STS can help clarify the TS intent.

lnservice Testing Pursuant to 10 CFR 50.54, "Conditions of licenses," the applicable requirements of 10 CFR 50.55a are conditions of every nuclear power reactor operating license issued under 10 CFR Part 50. These requirements include inservice testing of pumps and valves at nuclear power reactors in accordance with the ASME OM Code as specified in 10 CFR 50.55a(f). The regulations in 10 CFR 50.55a(f) state, in part:

Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME BPV [Boiler and Pressure Vessel] Code and ASME Code for Operation and Maintenance of Nuclear Power Plants as specified in this paragraph. Each operating license for a boiling or pressurized water-cooled nuclear facility is subject to the following conditions

[referring to 10 CFR 50.55a(f)(1) through (f)(6)] ....

The ASME OM Code is a consensus standard, which is incorporated by reference into 10 CFR 50.55a. During the incorporation process, the NRC staff reviewed the ASME OM Code requirements for technical sufficiency and found that the ASME OM Code inservice testing program requirements were suitable for incorporation into the NRC's rules.

The regulation in 10 CFR 50.55(a)(f)(5)(ii) states, in part: "If a revised inservice test program for a facility conflicts with the technical specifications for the facility, the licensee must apply to the Commission for amendment of the technical specifications to conform the technical specifications to the revised program."

NUREG-1482, Revision 2, "Guidelines for lnservice Testing at Nuclear Power Plants: lnservice Testing of Pumps and Valves and lnservice Examination and Testing of Dynamic Restraints (Snubbers) at Nuclear Power Plants," Final Report, October 2013 (ADAMS Accession No. ML13295A020), provides guidance for the inservice testing of pumps and valves.

NUREG-0800, Section 3.9.6, "Functional Design, Qualification, and lnservice Testing Programs for Pumps, Valves, and Dynamic Restraints," Revision 3, March 2007 (ADAMS Accession No. ML070720041 ), provides guidance and acceptance criteria for the NRC staff review of the inservice testing program for pumps and valves.

3.0 TECHNICAL EVALUATION

The NRC staff evaluated the licensee's application, as supplemented, to determine if the proposed changes are consistent with the guidance, regulations, and licensing information discussed in Section 2.3 of this safety evaluation. In determining whether an amendment to a license will be issued, the Commission is guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. Among the considerations are whether the TSs, as amended, would provide the necessary administrative controls per 10 CFR 50.36(c)(5) (i.e., provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner). In making its determination as to whether to amend the license, the staff considered those regulatory requirements that are automatically conditions of the license through 10 CFR 50.54. Where the regulations already condition the license, there is no need for a duplicative requirement in the TSs; the regulations provide the necessary reasonable assurance of the health and safety of the public.

3.1 Deletion of the lnservice Testing Program from the TSs TS 4.2 specifies, in part, requirements for the TMl-1 lnservice Testing Program. TS 4.2.2 requires that ASME Code Class 1, 2, and 3 pumps and valves be tested in accordance with the ASME OM Code and applicable Addenda, as required by 10 CFR 50.55a, unless written relief from these requirements has been granted by the NRC. Through 10 CFR 50.54, the applicable requirements of 10 CFR 50.55a are conditions of every nuclear power reactor operating license issued under 10 CFR Part 50. These requirements include 10 CFR 50.55a(f), which specifies the requirements for the inservice testing of pumps and valves. Therefore, requiring the licensee to have an inservice testing program in TSs is duplicative of the license condition in 10 CFR 50.54. Thus, with the proposed TS changes, the licensee will still be required to maintain an inservice testing program in accordance with the ASME OM Code, as specified in 10 CFR 50.55a(f).

The NRC staff determined that the lnservice Testing Program requirements currently in TS 4.2 are not necessary to assure: (1) operation of the facility in a safe manner, or (2) 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. Based on this evaluation, the staff concludes that deletion of the lnservice Testing Program requirements from TS 4.2 are acceptable because the current requirements in TS 4.2 are not required by 10 CFR 50.36(c)(5) or 10 CFR 50.36(c)(3).

3.2 Definition of INSERVICE TESTING PROGRAM and Revisions to SRs The licensee proposes to revise the TS Definitions section to add Definition 1.27, "INSERVICE TESTING PROGRAM," with the following definition: "The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f)." The proposed definition of the INSERVICE TESTING PROGRAM is consistent with the definition in TSTF-545, Revision 3. The definition is acceptable to the NRC staff because it correctly refers to the inservice testing requirements in 10 CFR 50.55a(f).

The licensee also requested that all existing references to the "lnservice Testing Program" in SRs (TS Section 4) be revised to "INSERVICE TESTING PROGRAM," to reference the new TS defined term in lieu of the deleted program. The proposed change is consistent with the intent of TSTF-545, Revision 3, to replace the current references in SRs with the new definition. The NRC staff verified that for each SR reference to the "lnservice Testing Program," the licensee proposed to change the reference to "INSERVICE TESTING PROGRAM." The proposed change does not alter how the SR testing is performed. Based on its review, the staff determined that revising the SRs to refer to the new definition is acceptable because these SRs will continue to be performed in accordance with the requirements of 10 CFR 50.55a(f). The staff also determined that, with the proposed changes, 10 CFR 50.36(c)(3) will continue to be met because the SRs will continue to ensure 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.

3.3 Deviations from TSTF-545 In its application, the licensee identified the following deviations from TSTF-545, Revision 3:

1. TSTF-545, Revision 3, completely deletes the lnservice Testing Program from the TSs and renumbers the subsequent TS programs. The licensee proposes to delete the content of TS 4.2.2, but retains the TS number, and adds the word "DELETED." The licensee did not propose to renumber any TSs.

2. TMl-1 has custom TSs that differ in numbering, titles, and SRs from the improved STS on which TSTF-545, Revision 3, was based. The lnservice Testing Program is in TMl-1 TS 4.2 instead of TS Section 5.0. In addition, many of the SRs identified in TSTF-545, Revision 3, are not applicable to TMl-1.

The NRC staff finds that the proposed deviations are editorial in nature and the licensee's proposed TS changes remain consistent with the intent of TSTF-545, Revision 3. Therefore, the staff finds that the licensee's proposed TS changes are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Pennsylvania State official was notified of the proposed issuance of the amendment on March 16, 2017. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes SRs.

The NRC staff has determined that the amendment involves 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, which was published in the FR on November 8, 2016 (81 FR 78648), that the amendment involves no significant hazards consideration, and there has been no public comment on such finding. Accordingly, the amendment meets 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 amendment.

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 amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: Blake Purnell, NRR Date of issuance: May 2 6, 2o1 7

ML17073A067 *via e-mail OFFICE DORL/LPL3/PM DORL/LPL3/LA STSB/BC EPNB/BC NAME BPurnell SRohrer AKlein DAiiey*

DATE 04/04/2017 04/04/2017 04/05/2017 03/17/2017 OFFICE OGC NLO DORL/LPL3/BC DORL/LPL3/PM NAME AGhosh DWrona BPurnell DATE 04/19/2017 05/26/2017 05/26/2017