Text

R. E. Ginna - Issuance of Amendment Nos. 216, 216, 220, 220, and 143 - Adoption of TSTF-567, Rev. 1, Add Containment Sump Technical Specifications to Address GSI-191 Issues
	ML20167A007
Person / Time
Site:	Byron, Braidwood, Ginna
Issue date:	09/11/2020
From:	Joel Wiebe; Plant Licensing Branch III
To:	Bryan Hanson; Exelon Generation Co, Exelon Nuclear
	Wiebe J
References
	EPID L-2020-LLA-0222, GSI-191, TSTF-567, Rev 1
	Download: ML20167A007 (53)
	v • d • e

September 11, 2020 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, AND R. E. GINNA NUCLEAR POWER PLANT - ISSUANCE OF AMENDMENTS NOS. 216, 216, 220, 220, AND 143, RESPECTIVELY REGARDING ADOPTION OF TSTF-567, REVISION 1, ADD CONTAINMENT SUMP TS TO ADDRESS [GENERIC SAFETY ISSUES] GSI-191 ISSUES (EPID L-2020-LLA-0022)

Dear Mr. Hanson:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 216 to Renewed Facility Operating License No. NPF-72 and Amendment No. 216 to Renewed Facility Operating License No. NPF-77 for the Braidwood Station, Units 1 and 2, respectively; Amendment No. 220 to Renewed Facility Operating License No. NPF-37 and Amendment No. 220 to Renewed Facility Operating License No. NPF-66 for the Byron Station, Unit Nos. 1 and 2, respectively; and Amendment No. 143 to Renewed Facility Operating License No. DPR-18 for the R. E. Ginna Nuclear Power Plant. The amendments are in response to your application dated February 6, 2020 (Agencywide Documents Access and Management System Accession No. ML20037A725).

These amendments revise the above nuclear power plants technical specifications to adopt Traveler Technical Specification Task Force (TSTF)-567, Add Containment Sump TS

[Technical Specifications] to Address GSl-191 Issues.

B. Hanson A copy of the Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commissions biweekly Federal Register notice.

Sincerely,

/RA/

Joel S. Wiebe, Senior Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-456, STN 50-457, STN 50-454, STN 50-455, and 50-244

Enclosures:

1. Amendment No. 216 to NPF-72

2. Amendment No. 216 to NPF-77

3. Amendment No. 220 to NPF-37

4. Amendment No. 220 to NPF-66

5. Amendment No. 143 to DPR-18

6. Safety Evaluation cc: Listserv

EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-456 BRAIDWOOD STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 216 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 February 6, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 Commissions regulations and all applicable requirements have been satisfied.

Enclosure 1

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. 216, 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.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 60 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Nancy L.

Nancy L. Salgado Date: 2020.09.11 13:39:47 Salgado -04'00' Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: September 11, 2020

EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-457 BRAIDWOOD STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 216 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 February 6, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 Commissions regulations and all applicable requirements have been satisfied.

Enclosure 2

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. 216, 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.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 60 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Nancy L.

Nancy L. Salgado Date: 2020.09.11 13:40:23 Salgado -04'00' Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: September 11, 2020

ATTACHMENT TO LICENSE AMENDMENT NOS. 216 AND 216 RENEWED FACILITY OPERATING LICENSE NOS. NPF-72 AND NPF-77 BRAIDWOOD STATION, UNITS 1 AND 2 DOCKET NOS. STN 50-456 AND STN 50-457 Replace the following pages of the Renewed Facility Operating Licenses and the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT License No. NPF-72 License No. NPF-72 Page 3 Page 3 License No. NPF-77 License No. NPF-77 Page 3 Page 3 TSs TSs Page 3.5.2 - 4 Page 3.5.2 - 4 Page 3.5.3 - 2 Page 3.5.3 - 2 Page 3.6.8 - 1 Page 3.6.8 - 1 Page 3.6.8 - 2 Page 3.6.8 - 3 Page 5.5 - 21 Page 5.5 - 21

(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 Commissions 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. 216 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 No. 216

(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 Commissions 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. 216 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 No. 216

ECCS-Operating 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2.3 Verify ECCS locations susceptible to gas In accordance accumulation are sufficiently 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 pump starts automatically In accordance on an actual or simulated 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,D SI System (Cold Leg)

BRAIDWOOD UNITS 1 & 2 3.5.2 4 Amendment216

ECCS-Shutdown 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 The following SRs are applicable for all In accordance equipment required to be OPERABLE: with applicable SRs SR 3.5.2.1 SR 3.5.2.7 SR 3.5.2.3 SR 3.5.2.4 BRAIDWOOD UNITS 1 & 2 3.5.3 2 Amendment 216

Containment Sump 3.6.8 3.6 CONTAINMENT SYSTEMS 3.6.8 Containment Sump LCO 3.6.8 Two containment sumps shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Initiate action to Immediately containment sump(s) mitigate containment inoperable due to accident generated and containment accident transported debris.

generated and transported debris AND exceeding the analyzed limits. A.2 Perform SR 3.4.13.1. Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months AND A.3 Restore the 90 days containment sump(s) to OPERABLE status.

BRAIDWOOD UNITS 1 & 2 3.6.8 1 Amendment 216

Containment Sump 3.6.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One or more B.1 ----------NOTES----------

containment sump(s) 1. Enter applicable inoperable for Conditions and Required reasons other than Actions of LCO 3.5.2, Condition A. "ECCS - Operating," and LCO 3.5.3, "ECCS -

Shutdown," for emergency core cooling trains made inoperable by the containment sump(s).

2. Enter applicable Conditions and Required Actions of LCO 3.6.6, "Containment Spray and Cooling Systems," for containment spray trains made inoperable by the containment sump(s).

Restore the containment 7 days sump(s) to OPERABLE status.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not AND met.

C.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months BRAIDWOOD UNITS 1 & 2 3.6.8 2 Amendment 216

Containment Sump 3.6.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.8.1 Verify, by visual inspection, the In accordance containment sump(s) do not show structural with the damage, abnormal corrosion, or debris Surveillance blockage. Frequency Control Program BRAIDWOOD UNITS 1 & 2 3.6.8 3 Amendment 216

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 Safety Function Determination Program (SFDP) (continued)

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. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.16 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Nuclear Energy Institute (NEI) Topical Report (TR) NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 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.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 42.8 psig for Unit 1 and 38.4 psig for Unit 2 The maximum allowable containment leakage rate, La, at Pa, shall be 0.20% of containment air weight per day.

Leakage Rate acceptance criteria are:

a. Containment leakage rate acceptance criterion is 1.0 La.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are

< 0.60 La for the Type B and C tests and < 0.75 La for Type A tests; and BRAIDWOOD UNITS 1 & 2 5.5 21 Amendment 216

EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-454 BYRON STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 220 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 February 6, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 Commissions regulations and all applicable requirements have been satisfied.

Enclosure 3

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. 220, 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. This license amendment is effective as of the date of its issuance and shall be implemented within 60 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Nancy L.

Nancy L. Salgado Date: 2020.09.11 13:42:26 Salgado -04'00' Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: September 11, 2020

EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-455 BYRON STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 220 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 February 6, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 Commissions regulations and all applicable requirements have been satisfied.

Enclosure 4

2. Accordingly, the renewed operating 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. 220, 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.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 60 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Nancy L.

Nancy L. Salgado Date: 2020.09.11 13:43:03 Salgado -04'00' Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: September 11, 2020

ATTACHMENT TO LICENSE AMENDMENT NOS. 220 AND 220 RENEWED FACILITY OPERATING LICENSE NOS. NPF-37 AND NPF-66 BYRON STATION, UNIT NOS. 1 AND 2 DOCKET NOS. STN 50-454 AND STN 50-455 Replace the following pages of the Renewed Facility Operating License and the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT License No. NPF-37 License No. NPF-37 Page 3 Page 3 License No. NPF-66 License No. NPF-66 Page 3 Page 3 TSs TSs Page 3.5.2 - 4 Page 3.5.2 - 4 Page 3.5.3 - 2 Page 3.5.3 - 2 Page 3.6.8 - 1 Page 3.6.8 - 1 Page 3.6.8 - 2 Page 3.6.8 - 3 Page 5.5 - 21 Page 5.5 - 21

(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. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commissions 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. 220 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. 220

(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 Commissions 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. 220, 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 No. 220

ECCS-Operating 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2.3 Verify ECCS locations susceptible to gas In accordance accumulation are sufficiently 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 pump starts automatically In accordance on an actual or simulated 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,D SI System (Cold Leg)

BYRON UNITS 1 & 2 3.5.2 4 Amendment 220

ECCS-Shutdown 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 The following SRs are applicable for all In accordance equipment required to be OPERABLE: with applicable SRs SR 3.5.2.1 SR 3.5.2.7 SR 3.5.2.3 SR 3.5.2.4 BYRON UNITS 1 & 2 3.5.3 2 Amendment 220

Containment Sump 3.6.8 3.6 CONTAINMENT SYSTEMS 3.6.8 Containment Sump LCO 3.6.8 Two containment sumps shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Initiate action to Immediately containment sump(s) mitigate containment inoperable due to accident generated and containment accident transported debris.

generated and transported debris AND exceeding the analyzed limits. A.2 Perform SR 3.4.13.1. Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months AND A.3 Restore the 90 days containment sump(s) to OPERABLE status.

BYRON UNITS 1 & 2 3.6.8 1 Amendment 220

Containment Sump 3.6.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One or more B.1 --------------NOTES-----------

containment 1. Enter applicable Conditions sump(s) and Required Actions of inoperable LCO 3.5.2, "ECCS -

for reasons Operating," and LCO 3.5.3, other than "ECCS - Shutdown," for Condition A. emergency core cooling trains made inoperable by the containment sump(s).

2. Enter applicable Conditions and Required Actions of LCO 3.6.6, "Containment Spray and Cooling Systems,"

for containment spray trains made inoperable by the containment sump(s).

Restore the containment 7 days sump(s) to OPERABLE status.

C. Required C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action and associated AND Completion Time not met. C.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months BYRON UNITS 1 & 2 3.6.8 2 Amendment 220

Containment Sump 3.6.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.8.1 Verify, by visual inspection, the In accordance containment sump(s) do not show structural with the damage, abnormal corrosion, or debris Surveillance blockage. Frequency Control Program BYRON UNITS 1 & 2 3.6.8 3 Amendment 220

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 Safety Function Determination Program (SFDP) (continued)

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. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.16 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Nuclear Energy Institute (NEI) Topical Report (TR) NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 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.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 42.8 psig for Unit 1 and 38.4 psig for Unit 2 The maximum allowable containment leakage rate, La, at Pa, shall be 0.20% of containment air weight per day.

Leakage Rate acceptance criteria are:

a. Containment leakage rate acceptance criterion is 1.0 La.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C tests and < 0.75 La for Type A tests; and BYRON UNITS 1 & 2 5.5 21 Amendment 220

EXELON GENERATION COMPANY, LLC DOCKET NO. 50-244 R. E. GINNA NUCLEAR POWER PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 143 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 (Exelon Generation, the licensee) dated February 6, 2020, 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.

Enclosure 5

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. 143, are hereby incorporated in the renewed license.

Exelon Generation 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 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Nancy L.

Nancy L. Salgado Date: 2020.09.11 13:44:16 Salgado -04'00' Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: September 11, 2020

ATTACHMENT TO LICENSE AMENDMENT NO. 143 RENEWED FACILITY OPERATING LICENSE NO. DPR-18 R. E. GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244 Replace the following pages of the Renewed Facility Operating License and the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT License No. DPR-18 License No. DPR-18 Page 3 Page 3 TSs TSs Page 3.5.2-3 Page 3.5.2-3 Page 3.5.3-2 Page 3.5.3-2 Page 3.6.7-1 Page 3.6.7-2 Page 3.6.7-3 Page 5.5-11 Page 5.5-11

(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&Es 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 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. 143, 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 licensees 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 R. E. Ginna Nuclear Power Plant Amendment No. 143

ECCS - MODES 1, 2, and 3 3.5.2 SURVEILLANCE FREQUENCY Verify each breaker or key switch, as applicable, for each In accordance SR 3.5.2.3 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 pumps developed head at the test flow In accordance point is greater than or equal to the required developed with the head. INSERVICE TESTING PROGRAM In accordance SR 3.5.2.5 Verify each ECCS automatic valve in the flow path that is 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 SR 3.5.2.6 Verify each ECCS pump starts automatically on an actual with the or simulated actuation signal.

Surveillance Frequency Control Program In accordance SR 3.5.2.7 Verify ECCS locations susceptible to gas accumulation are with the sufficiently filled with water.

Surveillance Frequency Control Program R.E. Ginna Nuclear Power Plant 3.5.2-3 Amendment No. 122, 124, 143

ECCS - MODE 4 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 - NOTE -

An RHR train may be considered OPERABLE during alignment and operation for decay heat removal, if capable of being manually realigned to the ECCS mode of operation.

SR 3.5.2.4 and SR 3.5.2.7 are applicable for all In accordance with equipment required to be OPERABLE. applicable SR R.E. Ginna Nuclear Power Plant 3.5.3-2 Amendment 118, 143

Containment Sump 3.6.7 3.6 CONTAINMENT SYSTEMS 3.6.7 Containment Sump LCO 3.6.7 The containment sump shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment sump A.1 Initiate action to mitigate Immediately inoperable due to containment accident containment accident generated and transported generated and debris.

transported debris exceeding the analyzed AND limits.

A.2 Perform SR 3.4.13.1. Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months AND A.3 Restore the containment 90 days sump to OPERABLE status.

R.E. Ginna Nuclear Power Plant 3.6.7-1 Amendment No. 143

Containment Sump 3.6.7 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Containment sump B.1 --------------NOTES-------------

inoperable for reasons 1. Enter applicable other than Condition A. Conditions and Required Actions of LCO 3.5.2, "ECCS -

MODES 1, 2, and 3,"

and LCO 3.5.3, "ECCS

- MODE 4," for emergency core cooling trains made inoperable by the containment sump.

Restore the containment 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months sump to OPERABLE status.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND C.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months R.E. Ginna Nuclear Power Plant 3.6.7-2 Amendment No. 143

Containment Sump 3.6.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.7.1 Verify, by visual inspection, the containment sump In accordance does not show structural damage, abnormal with the corrosion, or debris blockage. Surveillance Frequency Control Program R.E. Ginna Nuclear Power Plant 3.6.7-3 Amendment No. 143

Programs and Manuals 5.5

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

c. A required system redundant to the inoperable 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. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.15 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in NEI 94-01, Rev. 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," Revision 2-A, dated October 2008.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 60 psig.

The maximum allowable primary containment leakage rate, La, at Pa, shall be 0.2% of containment air weight per day.

Leakage Rate acceptance criteria are:

a. Containment leakage rate acceptance criterion is 1.0 La. During the first plant startup following testing in accordance with this program, the leakage rate acceptance criteria are 0.60 La for the Type B and Type C tests and 0.75 La for Type A tests;

b. Air lock testing acceptance criteria are:

1. For each air lock, overall leakage rate is 0.05 La when tested at Pa, and

2. For each door, leakage rate is 0.01 La when tested at Pa.

R.E. Ginna Nuclear Power Plant 5.5-11 Amendment 136, 143

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 216 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-72, AMENDMENT NO. 216 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-77, AMENDMENT NO. 220 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-37, AMENDMENT NO. 220 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-66, AND AMENDMENT NO. 143 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 R. E. GINNA NUCLEAR POWER PLANT DOCKET NOS. STN 50-456, STN 50-457, STN 50-454, STN 50-455, AND 50-244

1.0 INTRODUCTION

By application dated February 6, 2020, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20037A725), Exelon Generation Company, LLC (the licensee) submitted a license amendment request (LAR) for Braidwood Station Units 1 and 2 (Braidwood), Byron Station Unit Nos. 1 and 2 (Byron), and R.E. Ginna Nuclear Power Plant (Ginna).

The amendment would revise Byron and Braidwood Technical Specification (TS) 3.5.2, ECCS

[Emergency Core Cooling System] - Operating, TS 3.5.3, ECCS - Shutdown, and TS 5.5.15, Safety Function Determination Program (SFDP). The amendment would revise Ginna TS 3.5.2, ECCS -MODES 1, 2, and 3, TS 3.5.3, ECCS - MODE 4, and TS 5.5.14, Safety Function Determination Program (SFDP). The proposed changes would also add a new Byron, Braidwood, and Ginna TS, Containment Sump, to Section 3.6, Containment Systems. The proposed changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-567, Revision 1, Add Containment Sump TS to Address GSI [Generic Safety Issue]-191 Issues, dated August 2, 2017 (ADAMS Accession No. ML17214A813). The U.S. Nuclear Enclosure 7

Regulatory Commission (NRC or Commission) issued a final safety evaluation (SE) approving TSTF-567, Revision 1, July 3, 2018 (ADAMS Accession No. ML18116A606).

The licensee has proposed several variations from the TS changes described in TSTF-567.

The variations are described in Section 2.2.5 and evaluated in Section 3.5 of this SE.

2.0 REGULATORY EVALUATION

2.1 System Description The containment sump system consists of the containment drainage flow paths, any design features upstream of the containment sump that are credited in the containment debris analysis, the containment sump strainers, the pump suction trash racks, and the inlet to the ECCS and Containment Spray System (CSS) piping. Following an accident, water from the Reactor Coolant System (RCS) and the ECCS and water sprayed into the containment from the CSS collects in the sump.

2.1.1 ECCS - Operating (Modes 1, 2, and 3)

The function of the ECCS is to provide core cooling and negative reactivity to ensure the reactor core is protected after any of the following accidents:

a. Loss-of-coolant accident (LOCA), coolant leakage greater than the capability of the normal charging system,

b. Rod ejection accident,

c. Loss of secondary coolant accident, including uncontrolled steam release or loss of feedwater, and

d. Steam generator tube rupture.

TS 3.5.2 is applicable in Modes 1, 2, and 3, and requires that two independent ECCS trains be operable to ensure that sufficient ECCS flow is available, assuming a single failure affects either train.

TS 3.5.2 helps ensure the following acceptance criteria for the ECCS following a LOCA, established by Title 10 of the Code of Federal Regulations (10 CFR) 50.46, will be met:

a. Maximum fuel element cladding temperature is 2200 degrees Fahrenheit,

b. Maximum cladding oxidation is 0.17 times the total cladding thickness before oxidation,

c. Maximum hydrogen generation from a zirconium water reaction is 0.01 times the hypothetical amount generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react,

d. Core is maintained in a coolable geometry, and

e. Adequate long-term core cooling capability is maintained.

TS 3.5.2 also limits the potential for a post-trip return to power following a main steam line break event and ensures that containment temperature limits are met.

2.1.2 ECCS - Shutdown (Mode 4)

TS 3.5.3 is applicable in Mode 4 and requires one of the two independent (and redundant)

ECCS trains to be operable to ensure that sufficient ECCS flow is available to the core following a design-basis accident.

2.1.3 Safety Function Determination Program (SFDP)

TS 5.5.15 (Braidwood and Byron) and TS 5.5.14 (Ginna) establish the SFDP, which implements the requirements of LCO 3.0.6. The SFDP ensures that a loss of safety function is detected and that appropriate actions are taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported systems conditions and required actions.

2.2 Proposed Changes to the TSs The proposed changes would revise TS 3.5.2, ECCS - Operating (Braidwood and Byron)

ECCS - Modes 1, 2, and 3 (Ginna), TS 3.5.3, ECCS - Shutdown (Braidwood and Byron)

ECCS - Mode 4 (Ginna) and TS 5.5.15 (Braidwood and Byron) and TS 5.5.14 (Ginna), Safety Function Determination Program. The proposed changes would also add a new TS, Containment Sump to Section 3.6, Containment Systems. The proposed changes are described below.

2.2.1 Proposed Changes to ECCS - Operating (Modes 1, 2, and 3)

Braidwood and Byron TS 3.5.2 currently contains surveillance requirement (SR) 3.5.2.8, which requires the following at a frequency in accordance with the surveillance frequency control program (SFCP):

Verify, by visual inspection, each ECCS train containment sump suction inlet is not restricted by debris and suction inlet trash racks and screens show no evidence of structural distress or abnormal corrosion.

Ginna TS 3.5.2 currently contains SR 3.5.2.7, which requires the following at a frequency in accordance with the surveillance frequency control program (SFCP):

Verify, by visual inspection, each RHR containment sump suction inlet is not restricted by debris and the containment sump screen shows no evidence of structural distress or abnormal corrosion.

The licensee proposed to modify and move the Braidwood and Byron SR 3.5.2.8 and the Ginna SR 3.5.2.7 from TS 3.5.2 and include it in the new containment sump TS.

This change is evaluated in Section 3.1 of this SE.

2.2.2 Proposed Changes to ECCS - Shutdown (Mode 4)

TS 3.5.3 currently contains SR 3.5.3.1, which refers to applicable SRs under TS 3.5.2. One of those referenced in the Braidwood and Byron TS is SR 3.5.2.8, as described in Section 2.2.1 of this SE. Because the licensee proposed to modify and move SR 3.5.2.8 from TS 3.5.2 and include it in the new containment sump TS, the licensee also proposed to delete the reference to SR 3.5.2.8 in SR 3.5.3.1. One of those referenced in the Ginna TS is SR 3.5.2.8. As described in Section 2.2.1, the licensee proposed to move the existing SR 3.5.2.7. The licensee proposes to renumber the existing SR 3.5.2.8 to SR 3.5.2.7.

This change is evaluated in Section 3.2 of this SE.

2.2.3 Proposed Changes to Safety Function Determination Program (SFDP)

The licensee proposed to add the following sentence at the end of TS 5.5.15 (Braidwood and Byron) and TS 5.5.14 (Ginna):

When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

This change is evaluated in Section 3.3 of this SE.

2.2.4 Proposed Addition of a New Containment Sump TS The licensee proposed to add TS 3.6.1.8 (Braidwood and Byron) and TS 3.6.7 (Ginna) requiring containment sumps1 to be operable during Modes 1, 2, 3, and 4. Condition A specifies that if the containment sump is inoperable due to containment accident generated and transported debris exceeding the analyzed limits, then the licensee is required to: (1) initiate action to mitigate the containment accident generated and transported debris immediately, (2) perform SR 3.4.13.1 once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and (3) restore the containment sump to OPERABLE status within 90 days (Required Actions A.1, A.2, and A.3, respectively). SR 3.4.13.1 requires verification that the reactor coolant system (RCS) operational leakage is within limits by performance of an RCS water inventory balance.

Condition B specifies that if the containment sumps are inoperable for reasons other than Condition A, then the licensee is required to restore the containment sumps to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for Ginna, and 7 days for Braidwood and Byron (Required Action B.1).

Required Action B.1 is modified by two notes for Braidwood and Byron, which direct (1) entering the applicable conditions and required actions of LCO 3.5.2, ECCS Operating, and LCO 3.5.3, ECCS Shutdown, for ECCS trains made inoperable by the containment sump, and (2) entering the applicable conditions and required actions of LCO 3.6.6, Containment Spray and Cooling Systems, for containment spray and cooling system (CSS) trains made inoperable by the containment sump. Ginna only contains a single note that is associated with the ECCS.

Ginna does not have the note for its CSS because its CSS does not take suction directly from the sump.

1 Braidwood and Byron have two containment sumps, while Ginna has a single containment sump.

Condition C specifies that if required actions and associated CTs under Conditions A and B are not met, then the licensee is required to be in Mode 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and Mode 5 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (Required Actions C.1 and C.2, respectively).

The licensee proposed to modify and move SR 3.5.2.8 currently located in TS 3.5.2. The new SR 3.6.19.1 requires the licensee to verify, by visual inspection, that the containment sump does not show structural damage, abnormal corrosion, or debris blockage in accordance with the SFCP.

Braidwood and Byrons containment sump design includes more than one containment sump.

Each sump supplies one train of ECCS and one train of CSS. The sumps are independent from one another. The NRC staff considers the sumps part of a single support system because containment accident generated and transported debris issues that would render one sump inoperable would render all the sumps inoperable. The new containment sump TS proposed is applicable to plants that have one or more than one containment sump. Ginna has a single containment sump that provides recirculation to both trains of ECCS.

This change is evaluated in Section 3.4 of this SE.

2.2.5 Variations from TSTF-567, Revision 1 The licensee is proposing the following variations from the TS changes described in TSTF-567 or the applicable parts of the NRC staffs SE of TSTF-567. The licensee proposed that Braidwood and Byron have a different LCO time for the new TS LCO 3.6.8. The model traveler has an LCO for Action B of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (or in accordance with the risk-informed CT program). The licensee proposed that Braidwood and Byron have an LCO of 7 days for a single sump out-of-service due to reasons other than Condition A. The NRC staff determined that this is consistent with ECCS and CSS TS that allow the longer action time for a single train out-of-service since each sump is associated with its own train of ECCS.

Because the Ginna CSS does not take suction from the containment sump due to that plants design, the TS markups do not include the second note in the new sump TS 3.6.7, Action B.

The second note requires entry into the applicable conditions of LCO 3.6.6 for CSSs. The note requires entry into applicable CSS conditions and required actions due to loss of the supporting system (containment sump). The CSS does not take suction directly from the containment sump during recirculation. Instead of taking suction from the containment sump, the CSS takes suction from the residual heat removal (RHR) system, which takes suction from the sump.

Therefore, the TS actions required for ECCS account for the dependence of the CSS system on RHR at Ginna, and the second note is not applicable.

The variations discussed in this section are evaluated in Sections 3.4 and 3.5. The variations do not affect the applicability of TSTF-567 or the NRC staff's SE to the proposed LAR.

The Ginna TSs use a different name for TS 3.5.3. In the Standard Technical Specifications (STSs), TS 3.5.3 is named ECCS - Shutdown. In the Ginna TSs, TS 3.5.3 is named ECCS -

Mode 4. This difference does not affect the conclusions of TSTF-567 or the associated staff SE.

The licensee proposed to use numbering different from the STSs in the Braidwood, Byron, and Ginna TSs. Instead of the STS TS 3.6.19, the new containment sump TS is proposed to be Byron and Braidwood TS 3.6.8, and Ginna TS 3.6.7.

The licensee proposed to replace the SR for inspection of the sump with the new TS.

Consequently, the numbering is consistent with the STSs (SR 3.5.2.8 for Byron and Braidwood).

The NRC staff concludes that these are conforming changes that do not affect the implementation of the TS at Byron and Braidwood.

For Ginna, the licensee proposed to delete TS 3.5.2.7 (containment sump SR) and SR 3.5.2.8 for gas accumulation would be renumbered as 3.5.2.7. SR 3.5.2.8 for gas accumulation is referenced in a note in SR 3.5.3.1. In the SR 3.5.3.1 note, the reference to 3.5.2.8 is replaced with 3.5.2.7. The NRC staff concludes that these are conforming changes that do not affect the implementation of the TS at Ginna.

The TS number for the SFDP in TSTF-567 is 5.5.15, but the SFDP is required by Ginna TS 5.5.14 for this program. Byron and Braidwood are consistent with the STS numbering for the SFDP.

The numbering differences described above are appropriate to reflect the numbering of existing TSs and do not affect the applicability of TSTF-567 or the associated NRC staff SE to the proposed LAR.

2.3 Applicable Regulatory Requirements and Guidance 2.3.1 Technical Specification Requirements Regulation 10 CFR Section 50.36(a)(1) requires each applicant for a license authorizing operation of a utilization facility to include in the application proposed TSs. That regulation also states, in part, that [a] summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications.

The regulation at 10 CFR 50.36(b) requires:

Each license authorizing operation of a utilization facility will include technical specifications. The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to [10 CFR] 50.34 [Contents of applications; technical information]. The Commission may include such additional technical specifications as the Commission finds appropriate.

The categories of items required to be in the TSs are provided in 10 CFR 50.36(c). As required by 10 CFR 50.36(c)(2)(i), the TSs will include LCOs, which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. The regulation at 10 CFR 50.36(c)(2)(i) requires that when an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met.

The regulation at 10 CFR 50.36(c)(3) requires TSs to include SRs, which are requirements relating to test, calibration, or inspection, to assure that the necessary quality of systems and components is maintained, facility operation will be within safety limits, and that the LCOs will be met.

The regulation at 10 CFR 50.36(c)(5) requires TSs to include administrative controls, which 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.

2.3.2 Guidance The guidance that the NRC staff considered in its review of this LAR included the following:

NUREG-0800, Revision 3, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition, Chapter 16.0, Technical Specifications, dated March 2010 (ADAMS Accession No. ML100351425),

provides guidance on review of TSs.

NUREG-1431, Revision 4.0, Standard Technical Specifications, Westinghouse Plants, Volume 1, Specifications, and Volume 2, Bases, dated April 2012 (ADAMS Accession Nos. ML12100A222 and ML12100A228, respectively).

3.0 TECHNICAL EVALUATION

3.1 Proposed Change to TS 3.5.2, ECCS - Operating The licensee proposed to modify and move SR 3.5.2.8 (Braidwood and Byron) and SR 3.5.2.7 (Ginna) from TS 3.5.2 to the new containment sump TS. Therefore, the licensee proposed the deletion of SR 3.5.2.8 (Braidwood and Byron) and SR 3.5.2.7 (Ginna).

The new SR 3.6.8.1 (Braidwood and Byron) and SR 3.6.7.1 (Ginna) does not limit the visual inspection to the suction inlet, trash racks, and screens, as currently required by the TSs, but instead requires inspection of the entire containment sump system. The containment sump system consists of the containment drainage flow paths, any design features upstream of the containment sump that are credited in the containment debris analysis, the containment sump strainers (or screens), the pump suction trash racks, and the inlet to the ECCS and CSS piping.

The NRC staff concludes the proposed change is acceptable because the existing requirements are either unchanged or expanded and will continue to ensure the containment sump is unrestricted (i.e., unobstructed) and stays in proper operating condition. Therefore, the staff finds the proposed change meets the requirements of 10 CFR 50.36(c)(3) because it provides an SR to assure the necessary quality of systems and components are maintained, that facility operation will be within safety limits, and that the LCOs will be met.

3.2 Proposed Change to TS 3.5.3, ECCS - Shutdown The licensee proposed to delete the reference to SR 3.5.2.8 (Braidwood and Byron) and SR 3.5.2.7 (Ginna) in TS 3.5.3. The NRC staff concludes the proposed change is acceptable because SR 3.5.2.8 is modified and moved to the new containment sump TS. The existing SR on the containment sump is augmented (requiring inspection of additional sump components) and moved to the new specification, and a duplicative requirement to perform the SR in TS 3.5.3 is removed. The new specification retains or expands the existing requirements on the containment sump, and retains the actions to be taken when the containment sump is inoperable (with the exception of adding new actions to be taken when the containment sump is inoperable due to containment accident generated and transported debris exceeding the analyzed limits). The new actions provide the licensee time to evaluate and correct the

condition instead of requiring an immediate plant shutdown. Therefore, the NRC staff finds that the proposed change to TS 3.5.3 meets the requirements of 10 CFR 50.36(c)(3) because it provides SRs to assure the necessary quality of systems and components are maintained, that facility operation will be within safety limits, and the LCOs will be met.

3.3 Proposed Change to TS 5.5.15 (Braidwood and Byron) and TS 5.5.14 (Ginna), Safety Function Determination Program (SFDP)

LCO 3.0.6 states:

When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, an evaluation shall be performed in accordance with Specification 5.5.15, Safety Function Determination Program (SFDP). 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.

When a support systems Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.

As required by TS 5.5.15, when a loss of safety function is determined to exist, the SFDP requires entry into the appropriate conditions and required actions of the LCO in which the loss of safety function exists. As allowed by TS 3.0.6, when a loss of function is solely due to a single TS support system, the appropriate LCO is the LCO for that support system. When the loss of function is the result of multiple support systems, the appropriate LCO is the LCO for the supported systems.

The licensee proposed to add the following sentence to TS 5.5.15:

When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

The NRC staff finds that the proposed addition to TS 5.5.15 clarifies the intent of the allowance (i.e., not to enter the Conditions and Required Actions) provided by LCO 3.0.6 and the SFDP for single train support systems. The NRC staff concludes the proposed change is acceptable since the actions for the support system LCO adequately address the inoperability of that system. Therefore, as required by 10 CFR 50.36(c)(5), the proposed change continues to provide adequate administrative controls to assure operation of the facility in a safe manner.

3.4 Proposed Addition of Containment Sump TS 3.4.1 Evaluation of the New TS The licensee proposed to add a new TS 3.6.8 (Braidwood and Byron) and TS 3.6.7 (Ginna) to address operability requirements of the containment sump.

The containment sump supports the post-accident operation of the ECCS and CSS. However, current ECCS TSs only contain SRs related to the containment sump, and the TS does not specify required actions that specifically address an inoperable containment sump. If the containment sump (an ECCS and CSS support system) were found to be inoperable, those respective LCOs would not be met. In order to address concerns related to containment sump operability due to debris accumulation described in NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors, September 13, 2004 (ADAMS Accession No. ML042360586), the licensee proposed to add a new specification to address containment sump inoperability and create a condition for when the sump is inoperable due to analyzed containment accident generated and transported debris.

Based on its evaluation below, the NRC staff determined that the proposed TS satisfies the requirements of 10 CFR 50.36(c)(2)(i) because the LCO specifies the lowest functional capability or performance levels of equipment required for safe operation of the facility. There is reasonable assurance that the required actions to be taken when the LCO is not met can be conducted without endangering the health and safety of the public.

3.4.2 Evaluation of the Applicability The new TS requires the containment sump to be operable during Modes 1, 2, 3, and 4. The ECCS and CSS TSs currently are applicable during Modes 1, 2, 3, and 4.

The NRC staff finds the proposed applicability is acceptable because the applicability is consistent with the applicability of the ECCS and CSS TSs, which are supported by the containment sump system.

3.4.3 Evaluation of Condition A The licensee has analyzed the susceptibility of the ECCS and CSS to the adverse effects of post-accident debris blockage and operation with debris-laden fluids. The licensee has established limits on the allowable quantities of containment accident generated debris that could be transported to the containment sump based on its current plant configuration. In the current TSs, if unanalyzed debris sources are discovered inside containment, if errors are discovered in debris-related analyses, or if a previously unevaluated phenomenon that can affect containment sump performance is discovered, the containment sump and the supported ECCS and CSS may be determined inoperable and the TSs would require a plant shutdown with no time provided in the TSs for the licensee to further evaluate the condition.

In order to address this situation and to provide time to evaluate the condition, the licensee proposed Condition A, which is applicable when the containment sump is inoperable due to containment accident generated and transported debris exceeding the analyzed limits.

Under Condition A, Required Action A.1, mandates immediate action to be initiated to mitigate the condition. The licensee's proposed TS Bases for Required Action A.1 provided the following examples of mitigating actions:

Removing the debris source from containment or preventing the debris from being transported to the containment sump; Evaluating the debris source against the assumptions in the analysis; Deferring maintenance that would affect availability of the affected systems and other LOCA mitigating equipment; Deferring maintenance that would affect availability of primary defense-in-depth systems, such as containment coolers; Briefing operators on LOCA debris management actions; or Applying an alternative method to establish new limits.

The NRC staff finds the proposed Required Action A.1 and its completion time (CT) are acceptable because they place urgency on the initiation of the appropriate actions that could mitigate or reduce the impact of the identified conditions.

Concurrently, Required Action A.2 mandates SR 3.4.13.1 (verify RCS operational LEAKAGE is within limits by performance of the RCS water inventory balance) to be performed at an increased frequency of once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . An unexpected increase in RCS leakage could be indicative of an increased potential for an RCS pipe break, which could result in debris being generated and transported to the containment sump.

The NRC staff finds the proposed Required Action A.2 and its CT are acceptable because the more frequent monitoring allows operators to act in a timely fashion to minimize the potential for an RCS pipe break while the containment sump is inoperable.

Proposed Required Action A.3 requires the inoperable containment sump to be restored to operable status in 90 days. The NRC staff finds the proposed Required Action A.3 and its CT are acceptable because they provide a reasonable amount of time for the licensee to diagnose, plan and possibly mitigate the unanalyzed debris condition and prevent a loss of ECCS and CSS safety function. In addition, 90 days is adequate given the conservatisms in the containment debris analysis and the proposed compensatory actions required to be implemented immediately by Required Action A.1. As discussed later in this SE section, the new SR will also require visual inspection of the containment sump system (including the containment drainage flow paths, any design features upstream of the containment sump that are credited in the containment debris analysis, the containment sump strainers, the pump suction trash racks, and the inlet to the ECCS and CSS piping for evidence of structural degradation, potential for debris bypass, and presence of corrosion or debris blockage) to ensure no loose debris is present and there is no evidence of structural distress or abnormal corrosion.

Based on the above, the NRC staff concludes that CONDITION A and its REQUIRED ACTIONS are acceptable.

3.4.4 Evaluation of Condition B Condition B specifies the actions required when the containment sump is inoperable for reasons other than containment accident generated and transported debris exceeding the analyzed limits.

Required Action B.1 requires restoring the containment sump to operable status and is modified by two notes. These two notes direct entry into the conditions and required actions for the supported systems (ECCS and CSS) upon entering Required Action B.1. Since Required Action B.1 directs entry to the corresponding ECCS and CSS TS, these notes retain the existing TS actions for ECCS or CSS trains made inoperable by an inoperable containment sump for reasons other than containment accident generated and transported debris exceeding the analyzed limits.

As discussed in Section 2.2.5, the Ginna CSS does not take suction from the containment sump but receives its water supply from ECCS (RHR). Therefore, the proposed TS for Ginna does not include the second note regarding CSS for Action B.1. Since CSS takes suction from ECCS (RHR), the actions required for ECCS in TSs 3.5.2 and 3.5.3 account for the dependence of the CSS system on RHR at Ginna. Therefore, omission of the second note is acceptable.

The proposed CT for Required Action B.1 is 7 days for Braidwood and Byron. This CT is consistent with the less limiting CT for a single inoperable ECCS or CSS train so that the ECCS and CSS TS actions control the licensee's response.

The proposed CT for Required Action B.1 is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for Ginna. This CT is consistent with the less limiting CT for a single inoperable ECCS or CSS train so that the ECCS and CSS TS actions control the licensee's response.

The NRC staff finds the proposed change is acceptable since it continues to provide remedial actions for when the containment sump is inoperable for reasons other than Condition A and ensures safe operation of the plant. In addition, the proposed CT is acceptable since it provides a reasonable time for repairs, and there is a low probability of an accident occurring during this period that would require the use of the containment sump.

3.4.5 Evaluation of Condition C If operators are unable to restore the affected containment sump to operable status under Conditions A or B, Required Action C.1, requires the unit to be in Mode 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months followed by Mode 5 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , as required by Required Action C.2.

The NRC staff finds this proposed condition and its required actions are acceptable because the condition is consistent with the STS, and the required action requires operators to place the unit in a condition in which the LCO no longer applies. In addition, the proposed CTs allow a reasonable amount of time to decrease from full-power conditions to the required plant conditions in an orderly manner and without challenging plant systems.

3.4.6 Evaluation of the New SR The licensee proposed a new SR in the new containment sump TS. This SR is currently located in TS 3.5.2 and referred to in TS 3.5.3. The proposed numbering for this new SR be

SR 3.6.8.1 (Braidwood and Byron) and SR 3.6.7.1 (Ginna). The frequency of the new SR is in accordance with the SFCP.

The proposed SR requires verification by visual inspection, and that the containment sump does not show structural damage, abnormal corrosion, or debris blockage.

The new SR is stated in generic terms and expands the scope of the required visual inspection to include the entire containment sump system. The entire containment sump system consists of the containment drainage flow paths, the containment sump strainers (or screens), the pump suction trash racks, and the inlet to the ECCS and CSS piping.

The NRC staff finds the proposed SR is acceptable since it expands the scope of inspection of the original SR. In addition, the proposed frequency is acceptable since it is the same as that currently required by the TSs. Therefore, the NRC staff finds that, as required by 10 CFR 50.36(c)(3), the necessary quality of systems will be maintained, facility operation will be within safety limits, and that the LCOs will be met.

3.4.7 Evaluation of Changes to the TS Bases As required by 10 CFR 50.36(a), the licensee submitted TS Bases changes (that corresponded to the proposed TS changes) to provide the reasons for the proposed TSs. The licensee stated that the TS bases changes are consistent with the bases changes in TSTF-567.

The TS Bases are not part of the TSs and the NRC staff does not review and approve the TS Bases, however the NRC staff noted that the description of the sump in the TS bases does not include the strainer, which is an important component. Additionally, although the model traveler SR 3.6.19.1 bases includes strainers in the description, the licensees corresponding bases do not. However, the background section of the bases for each facility, state that the strainers are contained in the sump and that they limit the quantity of debris that enters the sump suction piping. The staff concluded that the bases provided by the licensee are not the same as those in the model, but that they do provide an acceptable description of the TS and the inspections required.

3.4.8 Conclusion Regarding Proposed Containment Sump TS The new containment sump TS retains and expands the existing TS requirements with the exception of the addition of Condition A. Condition A provides a condition for an inoperable containment sump due to containment accident generated and transported debris exceeding the analyzed limits.

The NRC staff reviewed the proposed changes against the regulations and concludes that, for the reasons discussed above, the changes continue to meet the requirements of 10 CFR 50.36(c)(2)(i) and 50.36(c)(3) and thus provide reasonable assurance that the revised TSs will continue to have the requisite requirements and controls to operate Braidwood, Byron, and Ginna safely. Therefore, the NRC staff concludes that the proposed TS changes are acceptable.

3.5 Variations As discussed in Section 2.2.5, Braidwood and Byron proposed a different LCO time for the new TS LCO 3.6.8. The model traveler has an LCO for Action B of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (or in accordance with

the risk-informed CT program). Braidwood and Byron propose an LCO of 7 days for a single sump out of service due to reasons other than Condition A. This is consistent with ECCS and CSS TS that allow the longer action time for a single train out of service since each sump is associated with its own train of ECCS. The single sump CT is the same as a single train of ECCS and/or CSS. Since each train of ECCS/CSS has its own sump the loss of the sump is equivalent to loss of a train of ECCS/CSS and assigning the same CT is acceptable.

Section 2.2.5 discussed that Ginna, due to the plant design, does not include the second note in the new sump TS 3.6.7. This is because the CSS does not take suction directly from the sump, but instead is supplied by the ECCS (RHR). If two trains of ECCS (RHR) are inoperable in Modes 1, 2, and 3, the required action in TS 3.5.2 is to enter LCO 3.0.3 immediately. In Mode 4 only one train of ECCS is required to be operable by TS 3.5.3. If the required RHR system is inoperable, the required action is to immediately initiate action to restore the required RHR train to operable. These actions are appropriate for the applicable conditions. Since the CSS does not take suction from the sump, the omission of the second note is appropriate for Ginna.

The numbering and naming differences discussed in Section 2.2.5 do not affect the applicability of TSTF-567 or the associated staff SE to the proposed LAR. Therefore, these minor editorial variations are acceptable.

The TS number for the safety function determination program in the model is 5.5.15. Ginna uses 5.5.14 for this program. Byron and Braidwood are consistent with the standard TS.

This numbering difference is editorial and does not affect the applicability of TSTF-567 or the associated NRC staff SE to the proposed LAR.

3.6 Technical Evaluation Conclusion

The NRC staff determined that the proposed TS changes meet the standards for TS in 10 CFR 50.36 and are acceptable. As required by 10 CFR 50.36(c)(2), the LCOs specify the lowest functional capability or performance levels of equipment required for safe operation of the facility. The proposed changes to the SR assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met, to therefore satisfy 10 CFR 50.36(c)(3). In addition, the proposed changes to the administrative controls include provisions to assure safe operation of the facility as required by 10 CFR 50.36(c)(5).

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Illinois and New York State officials were notified of the proposed issuance of the amendment on June 12, 2020. Both State officials had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant change in the types or significant increase in the amounts of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding

published in the Federal Register on May 5, 2020 (85 FR 26730). 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(s) will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: S. Smith, NRR/DSS Date of Issuance: September 11, 2020

ML20167A007 *via e-mail OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DSS/STSB/BC OGC NAME JWiebe SRohrer VCusumano* JMcManus DATE 6/17/20 6/16/20 5/15/20 8/27/20 OFFICE NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME NSalgado* JWiebe DATE 9/11/20 9/11/20

v t e Letter Sequence Approval
EPID:L-2020-LLA-0222, TSTF-567 (Approved, Closed)
Initiation Acceptance... Administration Questions Answers Results Approval Other:
MONTHYEAR2020-09-11 [Table View]

ML20167A007

Text

SUBJECT:

Dear Mr. Hanson:

Enclosures:

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1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

3.6 Technical Evaluation Conclusion

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

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