ML20149K698

Issuance of Amendment Nos. 215, 215, 219, and 219 Permanent Extension of Type a and Type C Leak Rate Test Frequencies

ML20149K698
Person / Time
Site:	Byron, Braidwood
Issue date:	09/10/2020
From:	Joel Wiebe NRC/NRR/DORL/LPL3
To:	Bryan Hanson Exelon Generation Co, Exelon Nuclear
Joel Wiebe-NRR/DORL 301-415-6606
References
EPID L-2019-LLA-0208
Download: ML20149K698 (51)

Text

September 10, 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, AND BYRON STATION, UNIT NOS. 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 215, 215, 219, AND 219 RE: PERMANENT EXTENSION OF TYPE A AND TYPE C CONTAINMENT LEAK RATE TEST FREQUENCIES (EPID L-2019-LLA-0208)

Dear Mr. Hanson:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 215 to Renewed Facility Operating License No. NPF-72 and Amendment No. 215 to Renewed Facility Operating License No. NPF-77 for Braidwood Station, Units 1 and 2, respectively, and Amendment No. 219 to Renewed Facility Operating License No. NPF-37 and Amendment No. 219 to Renewed Facility Operating License No. NPF-66 for Byron Station, Unit Nos. 1 and 2, respectively. The amendments are in response to your application dated September 24, 2019 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML19269C108).

The amendments revise Braidwood Station, Units 1 and 2, and Byron Station, Unit Nos. 1 and 2, Technical Specification 5.5.16 by replacing the existing reference with a reference to Nuclear Energy Institute (NEI) Topical Report NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR [Title 10 of the Code of Federal Regulations] Part 50, Appendix J, Revision 3-A (ADAMS Accession No. ML12221A202), and the conditions and limitations specified in NEI 94-01, Revision 2-A (ADAMS Accession No. ML100620847), as the documents used to implement the performance-based containment leakage testing program in accordance with Option B of 10 CFR Part 50, Appendix J.

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, and STN 50-455

Enclosures:

1. Amendment No. 215 to NPF-72

2. Amendment No. 215 to NPF-77

3. Amendment No. 219 to NPF-37

4. Amendment No. 219 to NPF-66

5. 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. 215 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 September 24, 2019, 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.

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. 215 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 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION 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 10, 2020 Nancy L.

Salgado Digitally signed by Nancy L. Salgado Date: 2020.09.10 12:49:51 -04'00'

EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-457 BRAIDWOOD STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 215 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 September 24, 2019, 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.

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. 215 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 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION 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 10, 2020 Nancy L.

Salgado Digitally signed by Nancy L. Salgado Date: 2020.09.10 12:50:22 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NOS. 215 AND 215 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 (TSs), 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 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. 215 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. 215

(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. 215 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. 215

Programs and Manuals 5.5 BRAIDWOOD UNITS 1 & 2 5.5 21 Amendment 215 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.

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

EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-454 BYRON STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 219 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 September 24, 2019, 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.

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. 219 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 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION 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 10, 2020 Nancy L.

Salgado Digitally signed by Nancy L. Salgado Date: 2020.09.10 12:51:10

-04'00'

EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-455 BYRON STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 219 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 September 24, 2019, 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.

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. 219, 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 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION 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 10, 2020 Nancy L.

Salgado Digitally signed by Nancy L. Salgado Date: 2020.09.10 12:51:49 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NOS. 219 AND 219 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 Licenses and the Appendix A, Technical Specifications (TSs), with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE 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 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.

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. 219 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. 219

(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. 219, 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. 219

Programs and Manuals 5.5 BYRON UNITS 1 & 2 5.5 21 Amendment 219 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.

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

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 215 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-72, AMENDMENT NO. 215 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-77, AMENDMENT NO. 219 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-37, AND AMENDMENT NO. 219 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-66 EXELON GENERATION COMPANY, LLC BRAIDWOOD STATION, UNITS 1 AND 2 BYRON STATION, UNIT NOS. 1 AND 2 DOCKET NOS. STN 50-456, STN 50-457, STN 50-454, AND STN 50-455

1.0 INTRODUCTION

By application dated September 24, 2019 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML19269C108), Exelon Generation Company, LLC (the licensee) submitted license amendment requests (LARs) for Braidwood Station, Units 1 and 2 (Braidwood), and Byron Station, Unit Nos. 1 and 2 (Byron). The proposed amendments would revise Technical Specification (TS) 5.5.16, Containment Leakage Rate Testing Program, by replacing the reference to Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program (ADAMS Accession No. ML003740058), with a reference to the Nuclear Energy Institute (NEI) topical report (TR) NEI 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR [Title 10 of the Code of Federal Regulations] Part 50, Appendix J (ADAMS Accession No. ML12221A202), and the limitations and conditions specified in NEI 94-01, Revision 2-A (ADAMS Accession No. ML100620847).

Specifically, the proposed amendments would:

Increase the existing Type A integrated leakage rate test (ILRT) program test interval from 10 years to 15 years in accordance with NEI 94-01, Revision 3-A, and the limitations and conditions specified in NEI 94-01, Revision 2-A.

Adopt an extension of the existing containment isolation valve leakage rate testing (Type C) interval from 60 months, as currently permitted by 10 CFR Part 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, Option B, to a maximum 75-months for Type C leakage rate testing of selected components, in accordance with NEI 94-01, Revision 3-A.

Adopt the use of the American National Standards Institute (ANSI) and American Nuclear Society (ANS) joint standard ANSI/ANS 56.8-2002, "Containment System Leakage Testing Requirements.1 Adopt a more conservative test interval extension of 9 months, for Type A and Type C leakage rate tests in accordance with NEI 94-01, Revision 3-A.

2.0 REGULATORY EVALUATION

2.1 Proposed TS Changes

The requirements for the Braidwood and Byron containment leakage rate testing programs are specified in TS 5.5.16. The first paragraph of the current Braidwood and Byron TS 5.5.16 states:

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 RG 1.163, September 1995 and NEI 94-01, Revision 0.

The proposed change would revise the first paragraph of TS 5.5.16 to state:

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.

2.2 Background

The overall integrity (structural and leak tight integrity) of the primary containment is verified by a Type A ILRT, and the integrity of the penetrations and isolation valves are verified by Type B and Type C local leak rate tests (LLRTs), as required by 10 CFR Part 50, Appendix J. These tests are performed to verify the essential leak tight characteristics of the containment structure at the design-basis accident pressure. The Type A test also provides a verification of structural integrity. The leakage rate testing requirements of 10 CFR Part 50, Appendix J, Option B (Type A, Type B, and Type C tests) and the containment inservice inspection (CISI) 1 Available at https://webstore.ansi.org/Standards/ANSI/ANSIANS562002R2016.

requirements mandated by 10 CFR 50.55a, Codes and standards, demonstrate the continued integrity of the containment during its service life.

Braidwood and Byron TS 5.5.16 currently states, in part, that [t]he peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 42.8 psig [pounds per square inch gauge] for Unit 1 and 38.4 psig for Unit 2, and that [t]he maximum allowable containment leakage rate, La, at Pa, shall be 0.20% of containment air weight per day. As required by 10 CFR Part 50, Appendix J, and TS 5.5.16, the Type A, Type B, and Type C test results must not exceed the La with margin. For each unit, the containment overall leakage rate acceptance criterion is less than or equal to 1.0 La.

2.3 Regulatory Requirements The licensee requested a change to the renewed facility operating licenses for Braidwood and Byron in accordance with 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit. Section 50.54(o) of 10 CFR requires that the primary reactor containments for water-cooled power reactors shall be subject to the requirements set forth in 10 CFR Part 50, Appendix J. These regulations specify containment leakage testing requirements including the types required to ensure the leak-tight integrity of the primary reactor containment and systems and components which penetrate the containment. In addition, 10 CFR Part 50, Appendix J discusses leakage rate acceptance criteria, test methodology, frequency of testing, and reporting requirements for each type of test.

Appendix J to 10 CFR Part 50 includes two options, Option A - Prescriptive Requirements, and Option B - Performance-Based Requirements, either of which can be used to meet the 10 CFR Part 50, Appendix J requirements. The adoption of the Option B performance-based containment leakage rate testing for Type A, Type B, and Type C testing does not alter the basic method by which Appendix J leakage rate testing is performed; however, it does alter the frequency at which Type A, Type B, and Type C containment leakage tests must be performed.

Under the performance-based option of 10 CFR 50, Appendix J, the frequency for Type A tests is based on the historical performance of the overall containment system as a barrier to fission product releases to reduce the risk from reactor accidents. The frequency for Type B and Type C tests is based on the safety significance and historical performance of each boundary and isolation valve to ensure the integrity of the overall containment system as a barrier to fission product releases to reduce the risk from reactor accidents.

Section V.B.3 of 10 CFR 50, Appendix J, Option B, requires the TSs to include, by general reference, the RG or other implementation document used by the licensee to develop a performance-based leakage-testing program. The submittal for TS revisions must also contain justification, including supporting analyses, if the licensee deviates from methods approved by the U.S. Nuclear Regulatory Commission (NRC or Commission) and endorsed in RG 1.163.

As required by 10 CFR 50, Appendix J, Option B,Section III, Type A tests are conducted at periodic intervals based on the performance history of the overall containment system to measure the overall integrated leakage rate. The leakage rate test results must not exceed the maximum allowable leakage (La) at design-basis loss-of-coolant accident (DBLOCA) pressure (Pa) with margin, as specified in the TSs. Option B also requires that a general visual inspection for structural deterioration of the accessible interior and exterior surfaces of the containment system be conducted prior to each Type A test and at a periodic interval between tests based on the performance of the containment system. A general visual inspection is necessary as structural deterioration of the surfaces of the containment system may affect the containments leak-tight integrity.

Type B and Type C tests are performed based on the safety significance and historical performance of each boundary and isolation valve to ensure integrity of the overall containment system as a barrier to fission product release.

Section 50.55a, Codes and standards, of 10 CFR contains the CISI program requirements, which, along with the requirements of 10 CFR Part 50, Appendix J, ensure the continued leak-tightness and structural integrity of the containment during its service life.

Paragraph (a)(1) of 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants, states, in part, that the licensee:

shall monitor the performance or condition of structures, systems, or components, against licensee-established goals, in a manner sufficient to provide reasonable assurance that these structures, systems, and components are capable of fulfilling their intended functions. These goals shall be established commensurate with safety and, where practical, take into account industrywide operating experience.

Braidwood and Byron adopted Option B of 10 CFR Part 50, Appendix J, for integrated (Type A) and local (Types B and C) leakage rate testing with Amendment Nos. 73 and 81, respectively (ADAMS Accession No. ML020870051).

The NRC staff also considered the following regulatory requirements in 10 CFR 50.36, Technical specifications, where the Commission established its regulatory requirements related to the contents of the TSs. Specifically, 10 CFR 50.36(a)(1) states that [e]ach applicant for a license authorizing operation of a production or utilization facility shall include in his application proposed technical specifications in accordance with the requirements of this section. In addition, 10 CFR 50.36(c)(5) 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.

2.4 Regulatory Guidance NEI 94-01, Revision 0 (ADAMS Accession No. ML11327A025), provides methods for complying with Option B of 10 CFR Part 50, Appendix J, and allows for the extension of the performance-based Type A test interval for up to 10 years, based upon two consecutive successful tests. NEI 94-01, Revision 0, was endorsed by the NRC in RG 1.163 with some conditions.

NEI 94-01, Revision 2 (ADAMS Accession No. ML072970206), incorporated the Regulatory Positions in RG 1.163 and added provisions for extending Type A test intervals up to 15 years.

This revision of NEI 94-01 was supported by Electric Power Research Institute (EPRI) TR No. 1009325, Revision 2, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, dated August 2007 (ADAMS Accession No. ML072970208). The EPRI report provides a generic assessment of the risks associated with permanently extending the ILRT interval to 15 years, and it provides a risk-informed methodology to be used to confirm the risk impact of the ILRT extension on a plant-specific basis. Probabilistic risk assessment (PRA) methods are used in combination with ILRT performance data and other considerations to justify the extension of the ILRT interval. This is consistent with the NRC guidance provided in RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (ADAMS Accession No. ML17317A256), and RG 1.177, An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications (ADAMS Accession No. ML100910008), to support changes to test intervals.

The NRC staffs review of both NEI 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2, is described in the NRC safety evaluation (SE) dated June 25, 2008 (ADAMS Accession No. ML081140105). The SE found that NEI 94-01, Revision 2, describes an acceptable approach for implementing the optional performance-based requirements of 10 CFR Part 50, Appendix J, Option B. The NRC staff concluded that NEI 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2, are acceptable for referencing by licensees proposing to amend their containment leakage rate testing TSs, subject to the conditions listed in Section 4.0 of the SE. The SE was incorporated into NEI 94-01, Revision 2, and subsequently issued as NEI 94-01, Revision 2-A, on November 19, 2008. In October 2008, EPRI Report No. 1009325, Revision 2-A, was published, which incorporated the applicable changes identified in the SE.2 NEI 94-01, Revision 3 (ADAMS Accession No. ML112920567), added guidance for extending Type C LLRT intervals beyond 60 months. In a June 8, 2012, SE (ADAMS Accession No. ML121030286), the NRC staff concluded that NEI 94-01, Revision 3, describes an acceptable approach for implementing the optional performance-based requirements of 10 CFR Part 50, Appendix J, and is acceptable for referencing by licensees proposing to amend their containment leakage rate testing TSs, subject to two conditions related to Type C testing.

The SE was incorporated into Revision 3 and subsequently issued as NEI 94-01, Revision 3-A, on July 31, 2012. By letter dated December 6, 2012 (ADAMS Accession No. ML12226A546),

the NRC staff stated that NEI 94-01, Revision 3, is acceptable with the conditions and limitations contained in its SE dated June 8, 2012.

3.0 TECHNICAL EVALUATION

3.1 Type A ILRT Braidwood and Byron TS 5.5.16 currently requires Types A, B, and C testing in accordance with RG 1.163, which endorses a methodology for complying with Option B. Since the licensees adoption of Option B, the performance leakage rates are calculated in accordance with Section 9.1.1 of NEI 94-01, Revision 0, for Type A testing. Tables 3.1-1 and 3.1-2 below list the past as-found (AF) Type A ILRT results for Braidwood, Units 1 and 2, respectively. Similarly, Tables 3.1-3 and 3.1-4 below list the past AF Type A ILRT results in weight percent per day (wt.%/day) for Byron, Unit Nos. 1 and 2, respectively.

2 EPRI Report No. 1009325, Revision 2-A, is also identified as EPRI Report No. 1018243. This report is publicly available and can be found at www.epri.com by typing 1018243 in the search box.

Table 3.1-1: Braidwood Unit 1 Type A ILRT History(4)

Test Date 95% UCL Leakage Rate (wt.%/day)

(test pressure)

(1)

Volume Level Corrections (wt.%/day)

Type B and C Penalties (wt.%/day)

Component Isolated During ILRT Performance Leakage Rate (wt.%/day)

Acceptance Criteria (wt.%/day) 5/21/1986 0.0333 (3)

(3)

(3)

(3) 0.075 2/2/1991 0.05286 (3)

(3)

(3)

(3) 0.075 11/8/1995 0.06324 (3)

(3)

(3)

(3) 0.075 10/5/1998 0.0605 (48.2 psia, Pa=47.8 psig) 0.000135 0.01010 0

0.0707 0.075 9/24/2013 0.0497 (43.7 psig, Pa=42.8 psig) 0.00141 0.0046 0

0.05571 0.150 Table 3.1-2: Braidwood Unit 2 Type A ILRT History(4)

Test Date 95% UCL Leakage Rate (wt.%/day)

(test pressure)

(1)

Volume Level Corrections (wt.%/day)

Type B and C Penalties (wt.%/day)

Component Isolated During ILRT Performance Leakage Rate (wt.%/day)

Acceptance Criteria (wt.%/day) 9/6/1987 0.0490 (3)

(3)

(3)

(3) 0.075 9/12/1991 0.05259 (3)

(3)

(3)

(3) 0.075 10/271994 0.04944 (3)

(3)

(3)

(3) 0.075 5/4/1999 0.0487 (47.3 psia, Pa=44.4 psig) 0.000098 0.01401 0

0.06278 0.075 5/17/2014 0.106747 (40.4 psig, Pa=38.4 psig)

-0.001 0.0026 0

0.108347 0.150 Table 3.1-3: Byron Unit No. 1 Type A ILRT History(5)

Test Date 95% UCL Leakage Rate (wt.%/day)

(test pressure)

(1)

Volume Level Corrections (wt.%/day)

Type B and C Penalties (wt.%/day)

Component Isolated During ILRT Performance Leakage Rate (wt.%/day)

Acceptance Criteria (wt.%/day) 9/2/1983 0.0058 (3)

(3)

(3)

(3) 0.075 9/9/1988 0.0458 (3)

(3)

(3)

(3) 0.075 9/12/1991 0.01843 (3)

(3)

(3)

(3) 0.075 2/14/1998 0.05959 (49.19

psia, Pa=47.8 psig) 0.00063 0.01020 0

0.07042 0.075 9/29/2012 0.076724 (43.39

psig, Pa=42.8 psig) 0.0001215 0.00448 0

0.08133 0.150 Table 3.1-4: Byron Unit No. 2 Type A ILRT History(5)

Test Date 95% UCL Leakage Rate (wt.%/day)

(test pressure)(1)

Volume Level Corrections (wt.%/day)

Type B and C Penalties (wt.%/day)

Component Isolated During ILRT (wt.%/day)

Performance Leakage Rate (wt.%/day)

Acceptance Criteria (wt.%/day) 7/2/1986 0.0253 (3)

(3)

(3)

(3) 0.075 9/7/1990 0.0706 (3)

(3)

(3)

(3) 0.075 9/9/1993 0.06666 (3)

(3)

(3)

(3) 0.075 11/9/1999 0.05832 (47.3 psia, Pa=44.4 psig) 0.0011 0.01311 0

0.07253 0.075 10/19/2014 0.06562 (39.97 psig, Pa=38.4 psig) 0.00599 0.00816 0

0.07977 0.150 Tables 3.1-1 to 3.1-4 Notes:

(1) The minimum allowed test pressure is Pa x 0.96 per ANSI 56.8-1994, Section 3.2.11 "Type A Test Pressure" (2) Per TS 5.5.16 (3) Only the data in the last two rows are used to meet NEI 94-01, Revision 3-A. See discussion below.

(4) Data source LAR Attachment 1a (5) Data source LAR Attachment 1b The licensee stated that the data represented in the last two rows of Tables 3.1-1 to 3.1-4 of this SE were obtained consistent with the definition of Performance Leakage Rate" as defined in Section 5.0 of NEI 94-01, Revisions 2-A and 3-A.

Section 9.1.2 of NEI 94-01, Revision 3-A, states, in part: The elapsed time between the first and the last tests in a series of consecutive passing tests used to determine performance shall be at least 24 months. Since the elapsed time between the last two Type A tests was 15 years, Section 9.1.2 of NEI 94-01, Revision 3-A, has been satisfied.

In its application, the licensee stated that:

The performance leakage rate is calculated as the sum of the Type A upper confidence limit (UCL) and as-left minimum pathway leakage rate (MNPLR) leakage rate for all Type B and Type C pathways that were in service, isolated, or not lined up in their test position (i.e., drained and vented to containment atmosphere) prior to performing the Type A test. In addition, leakage pathways that were isolated during performance of the test because of excessive leakage must be factored into the performance determination. The performance criterion for Type A tests is a performance leak rate of less than 1.0 La.

Since this is identical to the definition in NEI 94-01, Revision 3-A, Section 5.0, the calculation of performance leakage rate is acceptable.

For unit startup following completion of Type A testing, TS 5.5.16.a establishes a maximum limit of less than or equal to 0.75 La, which equals 0.075 (before 2006) or 0.150 (after 2006) percent of containment air weight per day. Both Braidwood and Byron containments were designed for a leakage rate, La, not to exceed 0.1 percent (before 2006) or 0.2 percent (after 2006) by weight of containment air per day at the calculated peak pressure, Pa. As shown in Tables 3.1-1 to 3.1-4 of this SE, ILRT results demonstrated ample margin between each AF 95-percent UCL and La.

The past five ILRT results for each unit have confirmed that the containment leakage rates are acceptable with respect to La. The last two Type A tests for each unit had AF test results of less than 1.0 La at the DBLOCA pressure (Pa) and the guidelines in NEI 94-01, Revisions 2-A and 3-A, regarding acceptable performance history, have been met. Because of these test results, the NRC staff concludes that the performance history of the Type A ILRTs for Braidwood and Byron provide reasonable assurance that containment overall leakage will be maintained below the design-basis leak rate, consistent with the requirements in TS 5.5.16, and will fulfill the requirements of 10 CFR Part 50, Appendix J, Option B, with a test interval of 15 y e a r s.

3.2 Type B and Type C Leak Rate Testing Program Type B testing ensures that the leakage rate of individual containment penetration components is acceptable. Type C testing ensures that individual containment isolation valves (CIVs) are essentially leak tight. In addition, aggregate Type B and Type C leakage rates support the leak tightness of both containments by minimizing potential leakage paths.

Braidwood, Units 1 and 2 In accordance with Braidwood, Units 1 and 2, TS 5.5.16, the allowable maximum pathway total Types B and C leakage is 0.6 La, which equates to 540.48 standard cubic feet per hour (scfh)for Unit 1 and 499.12 scfh for Unit 2.

As stated in the LAR, the Type B and Type C test results for Braidwood has shown substantial margin between the actual AF and as-left (AL) outage summations and the regulatory requirements as described below:

(1)

The actual AF minimum pathway leak rate average for Braidwood Unit 1 shows an average of 6.323 percent of 0.6 La with a high of 7.777 percent of 0.6 L a.

(2)

The actual AL maximum pathway leak rate average for Braidwood Unit 1 shows an average of 10.535% of 0.6 La with a high of 12.380% 0.6 La.

(3)

The actual AF minimum pathway leak rate average for Braidwood Unit 2 shows an average of 5.769% of 0.6 La with a high of 13.761% 0.6 L a.

(4)

The actual AL maximum pathway leak rate average for Braidwood Unit 2 shows an average of 9.129% of 0.6 La with a high of 11.467% 0.6 L a.

The NRC staff reviewed the Types B and C LLRT Combined As-Found/As-Left Trend Summary contained in Attachment 1a, Tables 3.5.5-1 to 3.5.5-2, of the LAR. Using the provided La values the NRC staff confirmed the accuracy of the licensees leakage rates calculations.

As stated in Attachment 1a, Section 3.5.5.1 Type B and Type C Local Leak Rate Test Program Implementation Review," of the LAR:

No Braidwood Station Units 1 and 2 Type B or Type C components on an extended frequency have exceeded their administrative limits over the last two refueling outages.

Based on the NRC staffs review of Attachment 1a, Tables 3.5.5-1 and 3.5.5-2, of the LAR, the aggregate results of the "As-Found Minimum Pathway" for all Braidwood Types B and C tests from 2009 through 2018 demonstrate a history of adequate maintenance since the aggregate test results at the end of each operating cycle were all well below (i.e., greater than 86 percent margin) the acceptance criteria of less than or equal to 0.6 La. Based on the review of the historical information, the NRC staff finds that the licensee is adequately implementing the requirements of its Appendix J, Option B, performance-based testing program.

The NRC staff finds that the Types B and C tests for both Braidwood units were less than the design-basis leak rate and meet the guidelines in NEI 94-01, Revisions 2-A and 3-A, regarding acceptable performance history. Therefore, the NRC staff concludes that the results of the Type B and Type C tests provide reasonable assurance that as-found minimum pathway totals will be maintained below the design-basis leak rate at Braidwood, Units 1 and 2, consistent with the requirements in TS 5.5.16 and fulfill the requirements of 10 CFR Part 50, Appendix J, Option B, when the Type C test interval is 75 months.

Byron, Unit Nos. 1 and 2 In accordance with Byron, Unit Nos. 1 and 2, TS 5.5.16, the allowable maximum pathway total Types B and C leakage is 0.6 L a. Prior to September 2014, for Unit No. 1, 1.0 L a was 899.03 scfh and 0.6 L a was 539.42 scfh, and for Unit No. 2 La was 829.99 scfh and 0.6 L a was 497.99 scfh. In September 2014, the L a was revised. For Unit No. 1, 1.0 La was revised to 815.75 scfh and 0.6 L a to 489.45 scfh and for Unit No. 2, 1.0 L a was revised to 753.97 scfh and 0.6 L a to 452.38 scfh.

As stated in the LAR, the Types B and C test results for Byron have shown substantial margin between the actual AF and AL outage summations and the regulatory requirements as described below:

(1) The actual AF minimum pathway leak rate average for Byron, Unit No. 1 shows an average of 11.05% of 0.6 La with a high of 18.16% 0.6 L a.

(2) The actual AL maximum pathway leak rate average for Byron, Unit No. 1 shows an average of 18.07% of 0.6 La with a high of 23.29% 0.6 L a.

(3) The actual AF minimum pathway leak rate average for Byron, Unit No. 2 shows an average of 14.67% of 0.6 La with a high of 19.29% 0.6 L a.

(4) The actual AL maximum pathway leak rate average for Byron, Unit No. 2 shows an average of 24.24% of 0.6 La with a high of 28.63% 0.6 L a.

The NRC staff reviewed the Types B and C LLRT Combined As-Found/As-Left Trend Summary contained in Attachment 1b, Tables 3.5.5-1 to 3.5.5-2, of the LAR. Using the provided La values, the NRC staff confirmed the accuracy of the licensees leakage rates calculations.

As stated in Attachment 1b, Section 3.5.5.1, Type B and Type C Local Leak Rate Testing Program Implementation Review, of the LAR, Byron Station Units 1 and 2 have no Type B components on the nominal test frequency of 30 months due to an as-found test failure.

Byron Station Unit 1 has two (2) Type C components on the nominal test frequency of 30 months due to an as found test failure. Unit 2 has no Type C component on the nominal test frequency of 30 months due to an as-found test failure.

Based on the review of Attachment 1b, Tables 3.5.5-1 and 3.5.5-2 of the LAR, the aggregate results of the As-Found Minimum Pathway for all Byron Type B and C tests from 2008 through 2019 demonstrate a history of adequate maintenance since the aggregate test results at the end of each operating cycle were all well below (i.e., greater than 81 percent margin) the Types B and C test leakage rate acceptance criteria of less than or equal to 0.6 L a. Based on the review of the historical information, the NRC staff finds that the licensee is adequately implementing the requirements of its Appendix J, Option B, performance-based testing program.

The NRC staff finds that the Types B and C tests for Byron, Unit Nos. 1 and 2, were less than the design-basis leak rate and meet the guidelines in NEI 94-01, Revisions 2-A and 3-A (which concern acceptable performance history). Therefore, the NRC staff concludes that the results of the Type B and Type C tests provide reasonable assurance that AF minimum pathway totals will be maintained below the design-basis leak rate at Byron Station, Unit Nos. 1 and 2 consistent with the requirements in TS 5.5.16 and will meet the requirements of 10 CFR Part 50, Appendix J, Option B, when the Type C test frequency is 75 months.

3.3 Containment Inspection and Testing Programs 3.3.1 Containment Inservice Inspection Program In Section 3.5.3 of the LAR (Attachment 1a for Braidwood and Attachment 1b for Byron), the licensee described its CISI program developed in accordance with the requirements incorporated by reference in 10 CFR 50.55a, Codes and standards. Visual inspections of the concrete are conducted in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code),Section XI, Subsection IWL, while inspections of the liner are conducted in accordance with ASME Code,Section XI, Subsection IWE, and the Containment Coatings Inspection and Assessment Program. The NRC staff reviewed the information provided on the containment inspection programs.

The licensee stated in its LAR that it is implementing its CISI program in accordance with Subsections IWE and IWL of the applicable edition and addenda of the ASME Code,Section XI.

The CISI plan for ASME Class MC (metal containment) and CC (concrete containment) components and structures for the Braidwood third 10-year CISI interval was developed in accordance with the requirements of 10 CFR 50.55a and the 2013 Edition of the ASME Code,Section XI, subject to the limitations and modifications contained within paragraph (b) of 10 CFR 50.55a. The CISI plan addresses Subsections IWE and IWL, Mandatory Appendices of ASME Code,Section XI, approved IWE Code Cases, and approved alternatives through relief requests and SEs. The Braidwood fourth ISI interval and third CISI interval are effective from August 29, 2018, through July 28, 2028, for Unit 1, and from November 5, 2018, through October 16, 2028, for Unit 2. For Byron, the fourth ISI interval and third CISI interval are effective from July 16, 2016, through July 15, 2025 for both units. Due to updates to the fourth ISI interval for Class 1, 2, and 3 components, and the third CISI interval for Class MC and CC components at Braidwood, all the ISI and CISI components and piping structural elements will be based on a common 2013 Edition ASME Code of Record. For Byron, the common ASME Code of Record for the fourth ISI interval and the third CISI interval is the 2007 Edition with the 2008 Addenda of the ASME Code,Section XI.

As the licensee discussed in Sections 3.1, Description of Containment System, of the LAR, the Braidwood and Byron containments are post-tensioned (PT) reinforced concrete structures with a carbon steel liner on the inside surface. Each containment consists of a cylindrical wall, a flat foundation mat, a shallow dome roof, and penetrations through the structure. The PT system consists of vertical and horizontal tendons in the cylinder wall and three-way tendons in the dome. The steel liner and its penetrations establish the leakage-limiting boundary of the containment and the PT reinforced concrete structures provide containment structural integrity.

The base foundation slab is conventionally reinforced with high-strength reinforcing steel, and a continuous access gallery is provided beneath the base slab for access to the vertical tendons.

Within each containment, the top of the base slab is lined with a steel liner plate to provide a leak tight membrane.

In Sections 3.7 of the LAR, Results of Containment Inspections, the licensee provided results for Braidwood and Byron containment inspections from 2011-2019, which also included primary containment coating condition assessments performed from 2016-2019. In general, the licensee identified light or moderate surface corrosion and rust and chipped and missing coating in various locations with no loss of material condition in both containments. The licensee stated that the conditions identified were reviewed by the licensees responsible individual and determined to have no impact on the containment structural integrity.

3.3.1.1 ASME Code,Section XI, Subsection IWL, Examinations ASME Code,Section XI, Subsection IWL, provides requirements for ISI of Class CC components and requires general visual examinations two times within a 10-year interval for concrete components. Similarly, concrete surfaces may be subject to detailed visual examination in accordance with item number L1.12 and paragraph IWL-2310(b) if declared to be Suspect Areas. ASME Code,Section XI, Subsection IWL, paragraph IWL-2330, requires that the responsible engineer be involved in the development, approval, and review of the CISI examinations.

In LAR Attachment 1a, Section 3.7.10, Units 1 and 2 IWL Examinations, 2016, the licensee discussed completion of the 30th year IWL visual examinations performed in 2016 for both Braidwood units. ASME Code Class CC containment concrete surfaces were examined in accordance with ASME Code,Section XI (2001 Edition/2003 Addenda), and additional criteria as specified in 10 CFR 50.55a, paragraphs (b)(2)(viii)(E) through (b)(2)(viii)(G). LAR Tables 3.7.10-1 and 3.7.10-2, Units 1 and 2 Concrete Indications-2016, provide discussion and evaluation of the conditions and indications identified during the general and detailed visual examinations performed at Braidwood in accordance with paragraphs IWL-2310 and IWL-2510 of the ASME Code,Section XI. The licensee stated that although the recoating of localized areas of the containment was performed in 2016, the recoating of the entire dome surfaces of both Braidwood containments was completed in 2018. The results of the examinations revealed no degradation that adversely affected the structural integrity of the containment structures. For Byron, Unit Nos. 1 and 2, LAR Section 3.7.12, Attachment 1b, discusses IWL examinations completed during the fourth 5-year ASME Code Class CC period (Spring 2016) for containment concrete surfaces. Both Byron containment structures are PT concrete designs inspected to the same requirements as both Braidwood units.

In LAR Attachment 1a, Sections 3.7.11/12, Tendon Surveillance Assessment for 2011/2017, the licensee reported the details of the 2011 and 2017 Braidwood, Units 1 and 2, 25th year (7th period) and 30th year containment structure PT system tendon surveillance, which assesses the quality and structural performance. In general, this surveillance consisted of a physical inspection of the system consisting of inspection for water, thread measurement, anchorage, and concrete, and was performed in accordance with the requirements of ASME Code,Section XI (2001 Edition with 2003 Addenda), Subsection IWL, as specified in 10 CFR 50.55a, Codes and standards. The licensee concluded that based on the data gathered during the 2011 ISI, no abnormal degradation of the PT system occurred; and for the 30th surveillance performed in 2017, the functional integrity of the selected PT met all the applicable code requirements. Section 3.7.14, Tendon Surveillance Assessment for August 2014, of b of the LAR, details the 30th year IWL Surveillance for Byron, Unit Nos. 1 and 2.

Based upon the evaluation of the ISI results, the licensee concluded that Byrons containment structures did not reveal abnormal degradation and that the PT system continues to meet the design requirements. The prestress loss trends indicate that both the Braidwood and Byron containment structures will continue to meet the design requirements beyond the projected 60-year life of the units.

3.3.1.2 ASME Code,Section XI, Subsection IWE Examinations In the LAR, the licensee described its ASME Code,Section XI, Subsection IWE, ISI program and provided a summary of recent inspections. Subsection IWE provides the rules and requirements for ISI of Class MC pressure-retaining surface areas subject to accelerated degradation and aging that require augmented examination in accordance with Examination Category E-C and paragraph IWE-1240 of the ASME Code,Section XI. Subsection IWE requires general visual examinations be performed three times within a 10-year interval for steel components, and paragraph IWE-2320 further requires that the responsible individual be involved in the development, performance, and review of the CISI examinations.

In LAR Attachment 1a, Section 3.7.5, Braidwood Unit 1 IWE Examination Refueling Outage (RFO) A1R19 (Fall 2016), the licensee stated that the general visual and augmented examinations were completed satisfactorily for CISI interval 2, period 3, RFO (refueling outage)

A1R19. The results of the visual examination of IWE surfaces for both ASME Class CC and MC components for Braidwood are detailed in LAR Table 3.7.5-2, Unit 1 ISI Containment Inspection Listing for RFO A1R19 (Fall 2016), and Table 3.7.7-3, Unit 2 ISI Containment Inspection Listing for RFO A2R19 (Spring 2017). Although some inactive corrosion was noted on the liner (e.g., as-found dry conditions, very small affected areas, inconsistent distribution pattern of the pitted areas), the licensee concluded that the containment liner is acceptable for full operation until the next CISI inspection period.

In LAR Attachment 1b, Section 3.7.6, Unit 1 IWE Examination, RFO B1R21, February 2017, for Byron, the licensee stated that the purpose of this inspection is to ensure that the structural integrity of ASME Class MC pressure retaining surfaces and Class CC metallic shell and penetration liners are maintained. All examinations were performed to comply with ASME Code,Section XI, Subsection IWE, 2007 Edition through the 2008 Addenda. The examination results were reviewed and compared to acceptance standards specified in IWE-3500 and evaluated by the appropriate personnel. The examination scope included inspection of the Byron containment liner seams and surfaces at the dome for any evidence of active degradation. The results of the visual examination of IWE surfaces for Byron were evaluated and found acceptable by the licensee and are presented in LAR Attachment 1b, Table 3.7.6-1, BYR Unit 1 IWE Examination RFO B1R21 (February 2017), and Table 3.7.9-1, BYR Unit 2 IWE Examination RFO B2R21 (April 2019). The licensee stated that the examination scope included, but was not limited to, inspection of the containment liner moisture barrier.

3.3.2 Coating Inspections In the LAR Section 3.5.1, Safety-Related (Service Level I) Protective Coatings Program, the licensee provided a discussion of the safety-related Service Level I Protective Coatings Program that is comparable to NRC RG 1.54, Service Level I, II, III, and In-Scope License Renewal Protective Coatings Applied to Nuclear Power Plants, Revision 3 (ADAMS Accession No. ML17031A288).

The licensees program provides a common approach in controlling, applying, maintaining, and periodically assessing Service Level I coatings used in areas inside the Braidwood and Byron reactor containments where the coating failure could adversely affect the operation of the emergency core cooling system ( ECCS) by clogging the suction strainers, which could possibly impair safe shutdown, and ensures qualified coatings are used inside primary containment and that those coatings are inspected and properly maintained. As stated in LAR Section 3.5.1.1, Unqualified/Degraded Coatings in Containment, the licensee has followed the guidance in the NRC SE related to generic safety issue (GSl)-191 (ML043280007) for determining the quantity of coating debris. The total amount of protective coating debris inside the containment building is limited to 45.9 cubic feet for each unit, which was considered acceptable based on ECCS suction strainer head loss testing. LAR Tables 3.5.1.1-1 through 3.5.1.1-6 summarize the estimated volumes of qualified and unqualified coatings inside the Braidwood and Byron primary containments and are below the 45.9 cubic feet limit.

In LAR Sections 3.7.1 through 3.7.4, the licensee discussed the results of recent primary containment coating condition assessments for Braidwood and Byron performed during RFOs from 2016-2019. As previously stated, Section 3.7, Results of Recent Containment Inspections, of Attachments 1a and 1b of the LAR, respectively discusses results for Braidwoods and Byrons primary containment coating condition assessments performed from 2016-2019. In general, the licensee identified areas of degraded coating that require repair that include light or moderate surface corrosion, areas of loose flaking coating and rust on the containment liner exposing the carbon steel substrate, areas of mechanical damage where the inorganic zinc primer has degraded, and chipped and missing coating in various locations for all four containments. No loss of material condition was identified in any of the four containments.

The licensee stated that a requirement associated with the containment safety-related coating evaluation is to identify areas on the liner plate and inner concrete walls where loose coating is to be scraped and collected. LAR Table 3.7.2.1-1 indicates the areas where loose coating was identified and removed. The licensees review of the identified conditions determined that there was no impact on the containment structural integrity or plant operations. In Section 3.8.1 of the LAR, the licensee stated that it will impose ASTM D5163-08, Standard Guide for Establishing a Program for Condition Assessment of Coating Service Level I Coating Systems in Nuclear Power Plants, requirements for Service Level I coatings condition assessment, reporting, evaluation, and documentation.

3.3.3 Conclusion Based on the above, and its review of the Braidwood and Byron CISI programs and recent inspection results, the NRC staff finds that the licensee is appropriately crediting the CISI inspections to meet the 10 CFR Part 50, Appendix J, visual inspection requirements. In addition, the staff finds that the licensee has an adequate CISI program in place as demonstrated by the implementation of overlapping inspection activities performed as part of the CISI programs and activities developed to support inspections of Service Level 1 protective coatings.

3.4 Conditions on NEI 94-01, Revision 2-A The Braidwood and Byron containments are subject to the requirements in 10 CFR Part 50, Appendix J, Option B, which allows test intervals for Types A, B, and C testing to be determined by using a performance-based approach. Currently, the licensee implements its Types A, B, and C testing in accordance with RG 1.163. The licensee proposed to revise TS 5.5.16 to require Types A, B, and C testing to be implemented in accordance with NEI 94-01, Revision 3-A, along with the conditions and limitations in NEI 94-01, Revision 2-A, to govern the test frequencies and the grace periods for the containment leakage rate tests.

The primary differences between Revision 2 and Revision 0 of NEI 94-01 are that Revision 2 incorporated the NRC conditions on Revision 0, specified in RG 1.163, and added provisions for extending Type A test intervals up to 15 years. In its SE dated June 25, 2008, the NRC staff concluded that the guidance in NEI 94-01, Revision 2, is acceptable for reference by licensees proposing to amend their containment leakage rate testing program TSs, subject to the six conditions listed in Section 4.1 of the SE. The licensee addressed these six conditions in its letter dated September 24, 2019, Attachment 1a (Braidwood) and Attachment 1b (Byron),

Section 3.9.1. The NRC staff evaluated these responses to determine whether the licensee adequately addressed these conditions.

3.4.1 NEI 94-01, Revision 2-A, Condition 1 NRC Condition 1 states:

For calculating the Type A leakage rate, the licensee should use the definition [of the performance leakage rate] in the NEI TR 94-01, Revision 2, in lieu of that in ANSI/ANS-56.8-2002.

The licensee stated that it will use the definition in Section 5.0 of NEI 94-01, Revision 3-A. The definition of the performance leakage rate in Revision 2, Revision 2-A, and Revision 3-A of NEI 94-01 are identical. Therefore, the NRC staff concludes that the licensee adequately addressed Condition 1.

3.4.2 NEI 94-01, Revision 2-A, Condition 2 NRC Condition 2 states:

The licensee submits a schedule of containment inspections to be performed prior to and between Type A tests.

NRC staff SE Section 3.1.1.3 reads, in part:

NEI TR 94-01, Revision 2, Section 9.2.3.2, states that: To provide continuing supplemental means of identifying potential containment degradation, a general visual examination of accessible interior and exterior surfaces of the containment for structural deterioration that may affect the containment leak-tight integrity must be conducted prior to each Type A test and during at least three other outages before the next Type A test if the interval for the Type A test has been extended to 15 years. NEI TR 94-01, Revision 2, recommends that these inspections be performed in conjunction or coordinated with the examinations required by ASME Code,Section XI, Subsections IWE and IWL. The NRC staff finds that these visual examination provisions, which are consistent with the provisions of regulatory position C.3 of RG 1.163, are acceptable considering the longer 15 year interval. Regulatory Position C.3 of RG 1.163 recommends that such examination be performed at least two more times in the period of 10 years. The NRC staff agrees that as the Type A test interval is changed to 15 years, the schedule of visual inspections should also be revised. Section 9.2.3.2 in NEI TR 94-01, Revision 2, addresses the supplemental inspection requirements that are acceptable to the NRC staff.

NEI 94-01, Revision 2-A and Revision 3-A, Section 9.2.3.2, Supplemental Inspection Requirements, both read:

To provide continuing supplemental means of identifying potential containment degradation, a general visual examination of accessible interior and exterior surfaces of the containment for structural deterioration that may affect the containment leak-tight integrity must be conducted prior to each Type A test and during at least three other outages before the next Type A test if the interval for the Type A test has been extended to 15 years. It is recommended that these inspections be performed in conjunction or coordinated with the ASME Boiler and Pressure Vessel Code,Section XI, Subsection IWE/IWL required examinations.

NEI 94-01, Revision 3-A, Section 9.2.1, Pretest Inspection and Test Methodology, reads, in part:

Prior to initiating a Type A test, a visual examination shall be conducted of accessible interior and exterior surfaces of the containment system for structural problems that may affect either the containment structure leakage integrity or the performance of the Type A test. This inspection should be a general visual inspection of accessible interior and exterior surfaces of the primary containment and components. It is recommended that these inspections be performed in conjunction or coordinated with the ASME Boiler and Pressure Vessel Code,Section XI, Subsection IWE/IWL required examinations.

In LAR Attachment 1a, Section 3.5.3, Containment lnservice Inspection Plan, the licensee provided Tables 3.5.3-1 and 3.5.3-2 on the Braidwood Component Examination Interval/Period/Outage Matrix.

In LAR Attachment 1b, Section 3.5.3, Containment lnservice Inspection Plan, the licensee provided Tables 3.5.3-1, 3.5.3-2, and 3.5.3-3 on the Byron Component Examination Interval/Period/Outage Matrix.

From its review of these tables, the NRC staff concludes that, for both Braidwood and Byron, general visual ISI examinations of accessible interior and exterior surfaces of the containment are scheduled for completion prior to the next Type A test, with at least three other 100 percent ISI inspections completed in between successive Appendix J Type A tests.

The licensee states in LAR Attachment 1a and Attachment 1b, Section 3.5.3 that:

Paragraph IWA-2430(c)(1) of ASME Section XI allows an inspection interval to be extended as much as one year or reduced without restriction, and paragraph IWA-2430(d) allows an inspection interval to be extended when a unit is out of service continuously for six months or more. The extension may be taken for a period of time not to exceed the duration of the outage. [Tables present]

intervals, periods, and detailed notes regarding any current extensions being taken that apply to [Braidwoods and Byrons] Fourth ISI Interval and Third CISI Interval.

The inspection of ISI Class CC components[, surfaces,] and tendons for the Third CISI Interval shall be performed in accordance with Paragraphs IWL-2410 and IWL-2420. [The LAR tables present] the inspection schedules.

For Byron, the licensee stated in LAR Attachment 1b, Section 3.5.3 that:

No significant conditions were identified in the First CISI Interval; however, significant conditions were identified in the Second CISI Interval as requiring application of additional augmented examination requirements under Paragraph IWE-2420 or IWL-2310.

In the Second CISI Interval, metal loss in excess of 10% of the [Byron Unit No. 1] containment liner at two locations below the MB at Elev. 377 have been identified as augmented surface areas requiring successive examinations in accordance with Paragraph IWE-2420(b). The indications were evaluated and accepted but re-examination of these two locations in the next inspection period (First Period, Third CISI Interval) is required.

These surface areas have been categorized in accordance with Table IWE-2500-1, Examination Category E-C, Item Number E4.11, requiring detailed visual examinations (i.e., VT-1) of 100% of the identified surface area each inspection period until the areas examined remain essentially unchanged for the next inspection period. When/If such areas no longer require augmented examination in accordance with Paragraph IWE-2420(d), the examination requirements and associated extent and frequency of Examination Category E-A apply for the remainder of the interval.

Based on the above information, the NRC staff finds that the licensee has submitted a schedule of the containment inspections to be performed prior to and between Type A tests. Therefore, the NRC staff concludes that the licensee has addressed and satisfied Condition 2.

3.4.3 NEI 94-01, Revision 2-A, Condition 3 NRC Condition 3 states:

The licensee addresses the areas of the containment structure potentially subjected to degradation.

In its LAR, Attachment 1a and Attachment 1b, Section 3.5.3, the licensee stated, in part:

The [Braidwood/Byron] Third Interval Containment lnservice Inspection (CISI) Program Plan was developed in accordance with the requirements of 10 CFR 50.55a subject to the limitations and modifications contained within Paragraph (b) of the regulation. [For Braidwood,] [t]his plan has been developed to comply with the 2013 Edition of American Society of Mechanical Engineers Boiler and Pressure Vessel Code [2007 Edition, 2008 Addenda for Byron],Section XI, Rules for lnservice Inspection of Nuclear Power Plant Components, and implements the requirements of:

TS 3.6.1, Containment TS 5.5.6, Pre-Stressed Concrete Containment Tendon Surveillance Program ASME Section XI, Subsection IWE, Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Plants ASME Section XI, Subsection IWL, Requirements for Class CC Concrete Components of Light-Water-Cooled Plants The [ISI] Program Plan details the requirements for the examination and testing of ISI Class 1, 2, 3, MC, and CC pressure retaining components, supports, containment structures, metal liners, and post-tensioning systems at [Braidwood/Byron], and Common (0). Unit Common components are included in the Unit 1 sections, reports, and tables. The ISI Program Plan also includes CISI, Risk-Informed lnservice Inspection,

Augmented Examinations, and System Pressure Testing requirements imposed on or committed to by [Braidwood/Byron]. The ISI Program Plan is also credited as the existing program for [Braidwood/Byron] License Renewal Aging Management Programs (AMPs).

In its LAR, Attachment 1a and Attachment 1b, Section 3.5.3.5, the licensee stated that metal containment surface areas, subject to accelerated degradation and aging, require augmented examinations pursuant to ASME Code,Section XI, Table IWE-2500-1, requirements for Examination Category E-C and paragraph IWE-1240. Similarly, concrete surfaces may be subject to detailed visual examination in accordance with Item Number L 1.12 and paragraph IWL-2310(b), if declared to be Suspect Areas.

In its LAR, Attachment 1a and Attachment 1b, Section 3.5.1, Safety-Related (Service Level I)

Protective Coatings Program, the licensee stated, in part:

The failure of the [safety-related] Service Level I [protective] coatings could adversely affect the operation of the ECCS by clogging the ECCS suction strainers. Proper maintenance of the Service Level I coatings ensures that coating degradation will not impact the operability of the ECCS systems. The program includes a visual examination of all reasonably accessible Service Level I coatings inside containment during every refueling outage and includes assessment and repair for any condition that adversely affects the intended function of Service Level I coatings.

Service Level I coatings will prevent or minimize the loss of material due to corrosion, but these coatings are not credited for managing the effects of corrosion for the carbon steel containment liners and components at [Braidwood/Byron]. This program ensures only that the Service Level I coatings maintain adhesion so as not to affect the intended function of the ECCS suction strainers.

The NRC staffs SE, Section 3.1.3, for NEI 94-01, Revision 2, reads, in part:

In approving for Type A tests the one-time extension from 10 years to 15 years, the NRC staff has identified areas that need to be specifically addressed during the IWE and IWL inspections including a number of containment pressure-retaining boundary components (e.g., seals and gaskets of mechanical and electrical penetrations, bolting, penetration bellows) and a number of the accessible and inaccessible areas of the containment structures (e.g., moisture barriers, steel shells, and liners backed by concrete, inaccessible areas of ice condenser containments that are potentially subject to corrosion).

In its LAR, Attachment 1a and Attachment 1b, Section 3.6.3, the licensee cited review of NRC Information Notice (IN) 2004-09, Corrosion of Steel Containment and Containment Liner (ADAMS Accession No. ML041170030). The IN was issued to alert licensees to recent occurrences of corrosion in freestanding metallic containments and in liner plates of reinforced and pre-stressed concrete containments. The licensee stated that the containment structural concrete and liner at Braidwood and Byron are ASME Code components covered and monitored by the CISI program. Both units containment liners are examined each period in accordance with ASME Section XI, Subsection IWE requirements and augmented examinations are performed requiring visual examination and ultrasonic exams. In RFO A2R07, liner degradation was discovered in Braidwoods containment liner and the entire moisture barrier was replaced for both units during RFO A1R08 and RFO A2R08. Engineering evaluations have been performed on the degraded areas during several RFOs and corrective actions have been completed. The licensee stated that no further actions are required for Braidwood. For Byron, similar corrective actions were implemented including moisture barrier replacement, application of service level I coatings, and continued monitoring of suspect areas, which have prevented liner plate degradation below the acceptance criteria.

In its LAR, Attachment 1a and Attachment 1b, Section 3.6.4, the licensee stated that it reviewed IN 2010-12, Containment Liner Corrosion (ADAMS Accession No. ML100640449). The IN was issued by the NRC to alert plant operators to events that occurred at nuclear power plant sites where the steel liner of the containment building was corroded and degraded. The licensee evaluated the IN and concluded that although the issue was applicable to Braidwood and Byron programs, implementation of IWE examinations, as well as applicable 10 CFR Part 50, Appendix J, testing, assure containment liner integrity with no operability concerns identified. The licensee stated that periodic examinations are performed during RFOs on metallic containment liners in accordance with IWE requirements.

Inaccessible Areas/Augmented Examinations The programmatic requirements for Class CC applications for inaccessible areas are in 10 CFR 50.55a(b)(2)(viii)(E), Concrete containment examinations: Fifth provision, which states:

For Class CC applications, the applicant or licensee must evaluate the acceptability of inaccessible areas when conditions exist that could indicate the presence of or could result in degradation to such inaccessible areas. For each inaccessible area identified, the applicant or licensee must provide the following in the ISI Summary Report as required by IWA-6000:

(1) A description of the type and estimated extent of degradation, and the conditions that lead to the degradation; (2) An evaluation of each area, and the result of the evaluation; and (3) A description of necessary corrective actions.

The programmatic requirements for Class MC applications for inaccessible areas are in 10 CFR 50.55a(b)(2)(ix)(A), Metal containment examinations: First provision, which states:

For Class MC applications, the following apply to inaccessible areas.

(1) The applicant or licensee must evaluate the acceptability of inaccessible areas when conditions exist that could indicate the presence of or could result in degradation to such inaccessible areas.

(2) For each inaccessible area identified for evaluation, the applicant or licensee must provide the following in the ISI Summary Report as required by IWA-6000:

(i) A description of the type and estimated extent of degradation, and the conditions that lead to the degradation; (ii) An evaluation of each area, and the result of the evaluation; and (iii) A description of necessary corrective actions.

As required by 10 CFR 50.55a(b)(2)(viii)(H), the licensee must provide the applicable information specified in paragraphs (b)(2)(viii)(E)(1), (2), and (3) in the ISI Summary Report.

In its LAR, Attachment 1a, Section 3.5.3.5, the licensee stated that a periodic containment tendon water monitoring and grease sampling program is implemented at Braidwood on a 5-year frequency to monitor exposure to free water and moisture in conjunction with the ASME Code,Section XI, Subsection IWL, AMP. Braidwood has performed augmented inspections during the first, second, and third CISI intervals related to additional tendons beyond those selected for the ASME Code,Section XI, Subsection IWL, program and containment liner examinations below the moisture barrier in accordance with paragraphs IWE-1240 and IWE-2420(b) of the ASME Code,Section XI. In its LAR, Attachment 1b, Section 3.5.3.5, the licensee stated that significant conditions were identified regarding metal loss in excess of 10 percent of the Byron, Unit No. 1, containment liner during the second CISI interval requiring additional augmented examinations pursuant to paragraphs IWE-2420(b) or IWL-2310 of the ASME Code,Section XI. The indications were evaluated and accepted but re-examination of the affected locations is required and will be performed during the first period of the third CISI interval.

Based on the above, the NRC staff finds that the licensee has addressed areas of the containment structure that are potentially subjected to degradation and, therefore, concludes that the licensee has adequately addressed Condition 3.

3.4.4 NEI 94-01, Revision 2-A, Condition 4 NRC Condition 4 states:

The licensee addresses any tests and inspections performed following major modifications to the containment structure, as applicable.

In its LAR, Sections 3.1.5 and 3.7.13 to Attachment 1a and Sections 3.1.5 and 3.7.15 to b, the licensee noted that no major containment modifications have been performed since the last ILRTs in 2013 and 2014 for Braidwood, Units 1 and 2, respectively, and in 2012 and 2014 for Byron, Unit Nos. 1 and 2, respectively. Steam generator replacements for Braidwood, Unit 1, and Byron Unit No. 1 were completed in 1998 with ILRTs performed after completion of the replacements to verify containment integrity. There are no major modifications planned that would require the performance of a Type A test or structural integrity test (SIT). The steam generators for Braidwood Unit 2 and Byron Unit No. 2 have not been replaced, and there are no plans to replace them. There are no major modifications planned that would require the performance of a Type A test or SIT.

The NRC staffs SE, Section 3.1.4, Major and Minor Containment Repairs and Modifications, for NEI 94-01, Revision 2, states, in part:

Repairs and modifications that affect the containment leakage integrity require LLRT or short duration structural tests as appropriate to provide assurance of containment integrity following the modification or repair. This testing shall be performed prior to returning the containment to operation. Article IWE-5000 of the ASME Code,Section XI, Subsection IWE (up to the 2001 Edition and the 2003 Addenda), would require a Type A test after major repair or modifications to the containment. In general, the NRC staff considers the cutting of a large hole in the containment for replacement of steam generators or reactor vessel heads, replacement of large penetrations, as major repair or modifications to the containment structure.

This condition is intended to verify that any major modification or maintenance repair of the containment since the last ILRT has been appropriately accompanied by either a structural integrity test or ILRT, and that any plans for such major modification also include appropriate pressure testing. As stated in the licensees response to Condition 4 in the LAR, no major modifications are planned for the Braidwood and Byron containment structures.

Based on the above, the NRC staff finds that the licensee has an adequate CISI program in place at Braidwood and Byron, as demonstrated by the implementation of overlapping inspection activities performed as part of the ASME Code,Section XI, Subsection IWE/IWL, programs and inspections performed as part of the Service Level 1 protective coatings and Maintenance Rule Structures Monitoring Program. The NRC staff finds that the licensee has addressed major modifications to the containment structure and concludes that the licensee has adequately addressed Condition 4.

The NRC staff also finds that the licensee is performing supplemental inspections to periodically examine and monitor aging degradation and, therefore, concludes that there is reasonable assurance that the containment structural and leak-tight integrity will continue to be maintained.

3.4.5 NEI 94-01, Revision 2-A, Condition 5 NRC Condition 5 states:

The normal Type A test interval should be less than 15 years. If a licensee has to utilize the provision of Section 9.1 of NEI TR 94-01, Revision 2, related to extending the ILRT interval beyond 15 years, the licensee must demonstrate to the NRC staff that it is an unforeseen emergent condition.

In its LAR, Attachment 1a, Table 3.9.1-1, the licensee stated:

[Braidwood] will follow the requirements of NEI 94-01, Revision 3-A, Section 9.1. This requirement has remained unchanged from Revision 2-A to Revision 3-A of NEI 94-01.

In accordance with the requirements of NEI 94-01, Revision 2-A, SE Section 3.1.1.2, [Braidwood] will also demonstrate to the NRC staff that an unforeseen emergent condition exists in the event an extension beyond the 15-year interval is required.

In its LAR, Attachment 1b, Table 3.9.1-1, the licensee stated:

[Byron] will follow the requirements of NEI 94-01, Revision 3-A, Section 9.1.

This requirement has remained unchanged from Revision 2-A to Revision 3-A of NEI 94-01.

In accordance with the requirements of NEI 94-01, Revision 2-A, SE Section 3.1.1.2, [Byron] will also demonstrate to the NRC staff that an unforeseen emergent condition exists in the event an extension beyond the 15-year interval is required.

The licensee stated that it will follow the requirements of NEI 94-01, Revision 3-A, Section 9.1, Introduction, which contains the relevant passage from the NRC staffs SE for NEI 94-01, Revision 2, Section 3.1.1.2, and states, in part:

Required surveillance intervals for recommended Type A testing given in this section may be extended by up to 9 months to accommodate unforeseen emergent conditions, but should not be used for routine scheduling and planning purposes. The NRC staff believes that extensions of the performance-based Type A test interval beyond the required 15 years should be infrequent and used only for compelling reasons. Therefore, if a licensee wants to use the provisions of Section 9.1 in TR NEI 94-01, Revision 2, the licensee will have to demonstrate to the NRC staff that an unforeseen emergent condition exists.

The NRC staff issued Regulatory Issue Summary (RIS) 2008-27, Staff Position on Extension of the Containment Type A Test Interval Beyond 15 Years Under Option B of Appendix J to 10 CFR Part 50 (ADAMS Accession No. ML080020394), which provides examples of justification for such an extension request. The above passage from SE Section 3.1.1.2 accurately reflects the NRC staff position.

Based on the above, the NRC staff finds that the licensee has demonstrated its understanding that any extension of the Type A test interval beyond the upper-bound performance-based limit of 15 years should be infrequent, and that any requested permission for such an extension will demonstrate to the NRC staff that an unforeseen emergent condition exists. Therefore, the NRC staff concludes that the licensee has addressed and satisfied NRC Condition 5.

3.4.6 NEI 94-01, Revision 2-A, Condition 6 Condition 6 applies only to plants licensed under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants. Condition 6 is not applicable to Braidwood and Byron because both are licensed under 10 CFR Part 50.

3.4.7 Conclusion Regarding the Conditions in NEI 94-01, Revision 2-A The NRC staff evaluated each of the six conditions in NEI 94-01, Revision 2-A, listed in Section 4.1 of the staffs June 25, 2008, SE and determined that the licensee adequately addressed conditions one through five and condition six is not applicable. Therefore, the NRC staff concludes it is acceptable for Braidwood and Byron to adopt the conditions and limitations in NEI 94-01, Revision 2-A, as part of the implementation documents listed in TS 5.5.16.

3.5 NEI 94-01, Revision 3-A The NRC published an SE with limitations and conditions for NEI 94-01, Revision 3, by letter dated June 8, 2012. In the SE, the NRC concluded that NEI 94-01, Revision 3, describes an acceptable approach for implementing the optional performance-based requirements of 10 CFR Part 50, Appendix J, and is acceptable for reference by licensees proposing to amend their containment leakage rate testing TSs, subject to two conditions. The SE was incorporated into NEI 94-01, Revision 3, and subsequently issued as NEI 94-01, Revision 3-A, on July 31, 2012. In its letter dated December 6, 2012 (ADAMS Accession No. ML12226A546), the NRC staff stated that NEI 94-01, Revision 3-A, is acceptable with the conditions and limitations contained in the NRC SE dated June 8, 2012.

3.5.1 NEI 94-01, Revision 3-A, Condition 1 NEI 94-01, Revision 3-A, Condition 1 presents the following three separate issues that are required to be addressed:

ISSUE 1 - The allowance of an extended interval for Type C LLRTs of 75 months carries the requirement that a licensees post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit.

ISSUE 2 - In addition, a corrective action plan shall be developed to restore the margin to an acceptable level.

ISSUE 3 - Use of the allowed 9-month extension for eligible Type C valves is only authorized for non-routine emergent conditions with exceptions as detailed in NEI 94-01, Revision 3-A, Section 10.1.

NEI 94-01, Revision 3-A, Condition 1 states, in part, that:

NEI TR 94-01, Revision 3, is requesting that the allowable extended interval for Type C LLRTs be increased to 75 months, with a permissible extension (for non-routine emergent conditions) of nine months (84 months total). The

[NRC] staff is allowing the extended interval for Type C LLRTs be increased to 75 months with the requirement that a licensee's post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit. In addition, a corrective action plan shall be developed to restore the margin to an acceptable level. The [NRC] staff is also allowing the non-routine emergent extension out to 84-months as applied to Type C valves at a site, with some exceptions that must be detailed in NEI 94-01, Revision 3. At no time shall an extension be allowed for Type C valves that are restricted categorically (e.g. BWR MSIVs

[boiling-water reactor main steam isolation valves]), and those valves with a history of leakage, or any valves held to either a less than maximum interval or to the base refueling cycle interval. Only non-routine emergent conditions allow an extension to 84 months.

In its LAR, Attachment 1a and Attachment 1b, Section 3.9.2, the licensee stated that the post-outage reports shall include the margin between the Type B and Type C minimum pathway leak rate (MNPLR) summation value, as adjusted to include the estimate of applicable Type C leakage understatement, and its regulatory limit of 0.6 La. When the potential leakage understatement adjusted Types B and C MNPLR total is greater than the Braidwood and Byron administrative leakage summation limit of 0.5 La, but less than the regulatory limit of 0.6 La, the licensee stated it shall perform an analysis and a corrective action plan shall be prepared to restore the leakage summation margin to less than the Braidwood and Byron leakage limit.

Additionally, the licensee stated that it will apply the 9-month allowable interval extension period only to eligible Type C components for non-routine emergent conditions.

The NRC staff compared the requirements of NEI 94-01, Revision 3-A, with the statements in the LAR and finds that the licensee intends to meet the requirements.

Therefore, the NRC staff concludes that the licensee addressed and satisfied Condition 1.

3.5.2 NEI 94-01, Revision 3-A, Condition 2 Condition 2 presents the following two separate issues that are required to be addressed:

ISSUE 1 - Extending the LLRT intervals beyond 5 years to a 75-month interval should be similarly conservative provided an estimate is made of the potential understatement and its acceptability determined as part of the trending specified in NEI TR 94-01, Revision 3-A, Section 12.1.

ISSUE 2 - When routinely scheduling any LLRT valve interval beyond 60 months and up to 75 months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Types B and C total and must be included in a licensees post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.

In its LAR, Attachment 1a and Attachment 1b, Section 3.9.2, the licensee stated that:

The change in going from a 60-month extended test interval for Type C tested components to a 75-month interval, as authorized under NEI 94-01, Revision 3-A, represents an increase of 25% in the LLRT periodicity. As such, [Braidwood/Byron] will conservatively apply a potential leakage understatement adjustment factor of 1.25 to the actual As-Left leak rate, which will increase the As-Left leakage total for each Type C component currently on greater than a 60-month test interval up to the 75-month extended test interval. This will result in a combined conservative Type C total for all 75-month LLRTs being carried forward and will be included whenever the total leakage summation is required to be updated (either while on line or following an outage).

When the potential leakage understatement adjusted leak rate total for those Type C components being tested on greater than a 60-month test interval up to the 75-month extended test interval is summed with the non-adjusted total of those Type C components being tested at less than or equal to a 60-month test interval, and the total of the Type B tested components results in the MNPLR being greater than the [Braidwood/Byron] administrative leakage summation limit of 0.50 La, but less than the regulatory limit of 0.6 La, then an analysis and corrective action plan shall be prepared to restore the leakage summation value to less than the [Braidwood/Byron] leakage limit. The corrective action plan shall focus on those components which have contributed the most to the increase in the leakage summation value and the manner of timely corrective action, as deemed appropriate, that best focuses on the prevention of future component leakage performance issues.

If the potential leakage understatement adjusted leak rate MNPLR is less than the [Braidwood/Byron] administrative leakage summation limit of 0.50 La, then the acceptability of the greater than a 60-month test interval up to the 75-month LLRT extension for all affected Type C components has been adequately demonstrated and the calculated local leak rate total represents the actual leakage potential of the penetrations.

NEI 94-01, Revision 3-A, also has a margin-related requirement as contained in Section 12.1, Report Requirements.

In its LAR, Attachment 1a and Attachment 1b, Section 3.9.2, the licensee stated that:

A post-outage report shall be prepared presenting results of the previous cycles Type B and Type C tests, and Type A, Type B and Type C tests, if performed during that outage... The report shall show that the applicable performance criteria are met and serve as a record that continuing performance is acceptable. The report shall also include the combined Type B and Type C leakage summation, and the margin between the Type B and Type C leakage rate summation and its regulatory limit. Adverse trends in the Type B and Type C leakage rate summation shall be identified in the report and a corrective action plan developed to restore the margin to an acceptable level.

Based on the above statements and programs implemented by the licensee, the NRC staff concludes that the licensee addressed and satisfied Condition 2.

3.5.3 Conclusion Regarding the Conditions in NEI 94-01, Revision 3-A The NRC staff evaluated the two conditions in NEI 94-01, Revision 2-A, and determined that the licensee adequately addressed each condition. Therefore, the NRC staff concludes it is acceptable for Braidwood and Byron to adopt the conditions and limitations in NEI 94-01, Revision 3-A, as part of the implementation documents listed in TS 5.5.16.

3.6 Probabilistic Risk Assessment of the Proposed Extension of the ILRT Test Intervals The licensee provided a plant-specific risk assessment for permanently extending the currently allowed containment Type A ILRT interval from 10 years to 15 years for Braidwood and Byron in LAR, Attachment 1a and 1b, Section 3.4.1 and in Attachment 3a and Attachment 3b.

The licensee stated that the plant-specific risk assessment follows the guidance in NEI 94-01; the methodology described in EPRI TR-1018243 (also identified as EPRI TR-1009325, Revision 2-A); and NRC RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, as well as the guidance in RG 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (ADAMS Accession No. ML090410014).

In addition, the licensee applied a methodology used at Calvert Cliffs Nuclear Power Plant to estimate the likelihood and risk implications of corrosion-induced leakage of steel liners going undetected during the extended test interval (ADAMS Accession No. ML020920100).

The licensee addressed each of the four conditions for the use of EPRI TR-1009325, Revision 2-A, which are listed in Section 4.2 of the NRC SE dated June 25, 2008. A summary of how each condition is met is provided in Sections 3.6.1 through 3.6.4 below.

3.6.1 Technical Adequacy of PRA The first condition stipulates that the licensee submit documentation indicating that the technical adequacy of its PRA is consistent with the requirements of RG 1.200 relevant to the ILRT extension LAR. This RG describes one acceptable approach for determining whether the technical adequacy of the PRA, in total or the parts, that are used to support an LAR is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision-making for light-water reactors.

Consistent with the information provided in Regulatory Issue Summary (RIS) 2007-06, Regulatory Guide 1.200 Implementation (ADAMS Accession No. ML070650428), the NRC staff used Revision 2 of RG 1.200 to assess technical adequacy of the PRA used to support this LAR. In Section 3.2.4.1 of the SE for NEI 94-01, Revision 2, and EPRI TR-1009325, Revision 2, the NRC staff states that Capability Category 1 of the ASME PRA standard shall be applied as the standard for assessing PRA quality for ILRT interval extensions, since approximate values of core damage frequency (CDF) and large early release frequency (LERF) and their distribution among release categories are sufficient to support the evaluation of changes to ILRT frequencies.

The licensee addressed the Braidwood PRA technical adequacy in LAR, Attachment 1a, Section 3.4.2, and Attachment 3a, Evaluation of Risk Significance of Permanent ILRT Extension. The licensee addressed the Byron PRA technical adequacy in LAR, Attachment 1b, Section 3.4.2, and Attachment 3b, Evaluation of Risk Significance of Permanent ILRT Extension. As discussed in Section 4 of Attachment 3a/3b of the LAR, the Braidwood and Byron risk assessments performed to support the LAR used the current respective Level 1 and Level 2 internal events PRA models of record which reflects the as-built, as-operated plants.

The 2013 versions of the Braidwood and Byron PRA models are the most recent risk profile evaluations for internal events. The licensee explained its approach to establishing and maintaining the technical adequacy and reflecting the as-built, as-operated plants in the PRA models. This approach included a PRA maintenance and update process, the use of self-assessments, and independent peer reviews.

The Braidwood PRA model for internal events received a full-scope industry peer review and self-assessment in July 2013. The findings and observations (F&O) closure was performed on February 2017 for the internal events and internal flooding models for Braidwood Units 1 and 2.

The model was reviewed against the current standard, ASME/ANS Ra-Sa-2009,3 Standard for Level l/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, and RG 1.200, Revision 2. The remaining set of open or partially resolved F&O from the independent review team assessment are described in Table A-1 in Attachment 3a of the LAR for internal events and fire PRAs with their impact on the LAR noted. There was one supporting requirement that was assessed as meeting Capability Categories 1, 2, and 3, but resulted in a finding. Finding SY-B12 is partially resolved, but a review of this finding found that there is minimal impact on this LAR.

The Byron PRA model for internal events received a full-scope peer review and self-assessment in July 2013. The F&O closure was performed on February 2017 for the internal events and internal flooding models for Byron Unit Nos. 1 and 2. The model was reviewed against the current standard, ASME/ANS Ra-Sa-2009, and RG 1.200, Revision 2. The remaining set of open or partially resolved findings and observations from the independent review team assessment are described in Table A-1 in Attachment 3b of the LAR for internal events and fire PRAs with their impact on the LAR noted. There was one supporting requirement that was assessed as meeting Capability Categories 1, 2, and 3, but resulted in a finding. Finding SY-B12 is partially resolved but a review of this finding found that there is minimal impact on this LAR.

With respect to external events, RG 1.174 stipulates that established acceptance guidelines are intended for comparison with a full-scope assessment of the change in the applicable risk metrics and recognizes that many PRAs are not full scope and PRA information of less than full scope may be acceptable. The methodology described in EPRI TR-1009325, Revision 2-A, which the NRC found satisfies the key principles of risk-informed decision-making of RG 1.174, discusses that if the external event analysis is not of sufficient quality or detail to allow direct application of the methodology, the quality or detail will be increased or a suitable estimate of the risk impact from the external events should be performed. This assessment can be taken from existing, previously submitted and approved analyses or another alternate method of assessing an order-of-magnitude estimate for contribution of the external event to the impact of the changed interval. Based on this, the licensee performed a bounding, order-of-magnitude, analysis of the potential impacts from external events. This analysis references the currently available information for external events models and information to develop an external events multiplier to be applied to the internal events results.

The Braidwood fire PRA self-assessment and full-scope peer review was performed in October 2015. The F&O closure was performed in February 2017 for models for Braidwood Units 1 and 2, using the NEI 07-12, Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines, fire PRA peer review process guidelines (ADAMS Accession No. ML083430464),

the ASME PRA standard, ASME/ANS RA-Sa-2009, and RG 1.200, Revision 2, as well as guidance from NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, (ADAMS Accession Nos. ML052580075 and ML052580118). The purpose of the review was to establish the technical adequacy of the fire PRA for the spectrum of potential risk-informed plant licensing applications for which the fire PRA may be used. The remaining set of open or partially resolved findings from the independent review team assessment are 3 Available from https://www.ans.org/store/.

described in Table A-1 of Attachment 3a of the LAR for the internal fire hazard group with their impact on this LAR noted. There were 12 supporting requirements that were assessed as Not Met, and two supporting requirements that were assessed as meeting Capability Category 1.

Also, there were three supporting requirements assessed as meeting Capability Category 1, 2, and 3 with associated findings. Findings FQ-E1-19-8, FQ-E1-19-9, FQ-F1-19-11, FQ-F1-19-15, HRA-D1-24-12, FQ-E1-25-9, FQ-F1-25-21, and FQ-F1-25-22 were determined to be documentation issues and did not impact the ILRT LAR. Findings CS-B1-16-4 and FSS-B2-20-8 were found to meet Capability Category I; however, the fire PRA model has been updated to include the necessary resolutions, as shown in calculation 54017-CALC-01 attached to the LAR. Findings UNC-A1-18-12 and UNC-A1-24 were resolved by updating files used in the uncertainty calculations. This resolution has been included in the fire PRA model update.

Findings FQ-E1 and PRM-B2 have been resolved through additional calculations and refinements and have been incorporated into the updated fire PRA model. Finding IGN-A7 was assessed as Capability Category 1, 2, and 3; however, the finding has been addressed via walkdowns of electrical cabinets and the data has been incorporated into the model. Therefore, a staff review of these findings and observations found that there is no material impact on this LAR.

The Byron fire PRA self-assessment and full-scope peer review was performed June 2015. The F&O closure was performed on February 2017 for models for Unit Nos. 1 and 2, using the NEI 07-12 fire PRA peer review process guidelines, the ASME PRA Standard, ASME/ANS RA-Sa-2009, and RG 1.200, Revision 2, as well as guidance from NUREG/CR-6850. The remaining set of open or partially resolved findings from the independent review team assessment are described in Table A-1 of Attachment 3b of the LAR for the internal fire hazard group with their impact on this LAR noted. There were 11 supporting requirements that were assessed as Not Met, and one supporting requirement that was assessed as meeting Capability Category 1. Also, there were three supporting requirements assessed as meeting Capability Category 1, 2, and 3 with associated findings. Findings FQ-E1-19-8, FQ-E1-19-9, FQ-F1-19-15, HRA-D1-24-12, FQ-E1-25-9, FQ-F1-25-21, and FQ-F1-25-22 were determined to be documentation issues and did not impact the ILRT LAR. Finding FSS-B2-20-8 was found to meet Capability Category 1; however, the fire PRA model has been updated to include the necessary resolutions as shown in calculation 54018-CALC-01 attached to the LAR. Findings UNC-A1-18-12 and UNC-A1-24 were resolved by updating files used in the uncertainty calculations. This resolution has been included in the PRA model update. Findings FQ-E1 and PRM-B2 have been resolved through additional calculations and refinements and have been incorporated into the updated fire PRA model. Finding IGN-A7 was assessed as Capability Category 1, 2, and 3; however, the finding has been addressed via walkdowns of electrical cabinets and the data has been incorporated into the model. Therefore, a staff review of these findings and observations found that there is no material impact on this LAR.

Based on review of the above information, the NRC staff finds that the licensee has addressed the relevant findings and gaps from the peer reviews and that they have no impact on the results of this LAR. Therefore, the NRC staff concludes that the internal events PRA model used by the licensee is of sufficient quality to support the evaluation of changes to ILRT frequencies. Accordingly, the NRC staff concludes that the first condition is met.

3.6.2 Estimated Risk Increase The second condition stipulates that the licensee submit documentation indicating that the estimated risk increase associated with permanently extending the ILRT interval to 15 years is small, consistent with the guidance in RG 1.174 and the clarification provided in the NRC SE for NEI 94-01, Revision 2-A, and EPRI TR-1009325, Revision 2-A. Specifically, a small increase in population dose should be defined as an increase in population dose of less than or equal to either 1.0 person-roentgen equivalent man (rem) per year or 1 percent of the total population dose, whichever is less restrictive. In addition, a small increase in conditional containment failure probability (CCFP) should be defined as a value marginally greater than that accepted in previous one-time 15-year ILRT extension requests. This would require that the increase in CCFP be less than or equal to 1.5 percentage points.

The NRC staff determined that the most relevant risk criterion for the proposed change is LERF because a change in the ILRT frequency does not impact CDF and Braidwood and Byron do not rely upon containment overpressure for ECCS performance (see Section 3.6.4 of this SE).

RG 1.174 considers a small change in LERF to be between 1E-7/year and 1E-6/year with a total LERF less than 1E-5. Thus, the associated risk metrics include LERF, delta LERF, population dose, CCFP, and delta CDF.

The licensee reported the results of the plant-specific risk assessment in LAR, Section 3.4 and in Attachments 3a/3b. External events are considered in Section 3.4 and Attachments 3a/3b and the impact of containment overpressure is assessed in Section 3.4. The reported risk impacts are based on a change in the Type A containment ILRT frequency from three tests in 10 years (the test frequency under 10 CFR Part 50, Appendix J, Option A) to one test in 15 years and account for the risk from undetected containment leaks due to steel liner corrosion. The following conclusions can be drawn from the licensees analysis associated with extending the Type A ILRT frequency:

1. RG 1.174 defines small changes in LERF to be between 1E-7/year and 1E-6/year with a total LERF less than 1E-5. RG 1.174 considers very small changes in LERF to be less than 1.0E-07/year. The increase in LERF resulting from a change in the Type A ILRT test frequency from 3-in-10 years to 1-in-15 years is estimated to be 4.04E-08/year and 3.98E-08/year for Braidwood, Units 1 and 2, respectively; and 5.48E-08/year and 4.87E-08/year for Byron, Unit Nos. 1 and 2, respectively, using the EPRI guidance with corrosion included. As such, the estimated change in LERF is determined to be very small using the acceptance guidelines of RG 1.174. When external event risk is included, the increase in LERF resulting from a change in the Type A ILRT test frequency from 3-in-10 years to 1-in-15 years is estimated as 5.48E-07/year and 5.54E-7/year for Braidwood, Units 1 and 2, respectively; and 6.15E-07/year and 6.24E-07/year for Byron, Unit Nos. 1 and 2, respectively, using the EPRI guidance and total estimated upper bound LERF is 7.34E-06/year and 7.13E-06/year for Braidwood, Units 1 and 2, respectively; and 7.33E-06/year and 7.80E-06/year for Byron, Unit Nos. 1 and 2, respectively. As such, the estimated change in LERF is determined to be small or between 1E-7/year and 1E-6/year with a total LERF less than 1E-5/year as specified in RG 1.174. The risk change resulting from a change in the Type A ILRT test frequency from 3-in-10 years to 1-in-15 years bounds the 1-in-10 years to 1-in-15 years risk change.

2. The effect resulting from changing the Type A test frequency to 1-in-15 years, measured as an increase in total population dose for those accident sequences influenced by Type A testing, for Braidwood is 0.135 person-rem/year for Unit 1 and 0.133 person-rem/year for Unit 2; and for Byron is 0.080 person-rem/year for Unit No. 1 and 0.071 person-rem/year for Unit No. 2. NEI 94-01 states that a small total population dose is defined as an increase of less than or equal to 1.0 person-rem/year, or less than or equal to 1 percent of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. The reported increase in total population dose is below the acceptance criteria provided in EPRI TR-1009325, Revision 2-A, and defined in Section 3.2.4.6 of the NRC SE for NEI 94-01, Revision 2-A.

Thus, the increase in the population dose risk for the proposed change is considered small and supportive of the proposed change.

3. The increase in the CCFP due to the change in test frequency to 1-in-15 years is reported for both Braidwood and Byron as 0.87 percent. NEI 94-01 states that increases in CCFP of less than or equal to 1.5 percentage points is small. This value is below the acceptance guidelines in Section 3.2.4.6 of the NRC SE for NEI 94-01, Revision 2-A, and supportive of the proposed change.

Based on the risk assessment results, the NRC staff concludes that the increase in LERF is small and consistent with the acceptance guidelines of RG 1.174, and that the increase in the total population dose and the magnitude of the change in the CCFP for the proposed change are small. The defense-in-depth philosophy is maintained because the independence of barriers will not be degraded because of the requested change, and the use of the quantitative risk metrics collectively ensures that the balance between prevention of core damage, prevention of containment failure, and consequence mitigation is preserved. Accordingly, the NRC staff concludes that the second condition is met.

3.6.3 Leak Rate for the Large Pre-Existing Containment Leak Rate Case The third condition stipulates that the average leak rate for the pre-existing containment large leak rate accident case (i.e., accident case 3b) used by licensees shall be 100 La instead of 35 La. As noted by the licensee in Section 4.0 of Attachment 3a and Attachment 3b of the LAR, the methodology in EPRI TR-1009325, Revision 2-A, incorporated the use of 100 La as the average leak rate for the pre-existing containment large leakage rate accident case (accident case 3b), and this value has been used in both the Braidwood and Byron plant-specific risk assessments. Accordingly, the NRC staff concludes that the third condition is met.

3.6.4 Applicability if Containment Over-Pressure is Credited for ECCS Performance The fourth condition stipulates that in instances where containment overpressure is relied upon for ECCS performance, an LAR is required to be submitted. In Section 3.2 of the LAR, the licensee states that the NPSH analysis for temperatures above 200 degrees Fahrenheit assumes that the vapor pressure of the recirculation sump liquid is equal to containment pressure. Thus, no reliance is placed on pressure and/or temperature transients to ensure adequate NPSH. Accordingly, the NRC staff concludes that the fourth condition is not applicable.

3.6.5 Conclusion Regarding Conditions on EPRI Report No. 1009325, Revision 2 The NRC staff evaluated each of the conditions on EPRI Report No. 1009325, Revision 2, listed in Section 4.2 of the NRC staffs SE dated June 25, 2008, and determined that each condition was adequately addressed in the LAR. These conditions are incorporated into NEI 94-01, Revision 2-A. Therefore, the NRC staff finds it acceptable for Braidwood and Byron to adopt the conditions and limitations specified in NEI 94-01, Revision 2-A, as part of the implementation documents listed in TS 5.5.16.

3.7 Technical Conclusion Based on the preceding regulatory and technical evaluations, the NRC staff finds that the licensee has adequately implemented its existing primary containment leakage rate testing program consisting of ILRTs and LLRTs. The results of the recent ILRTs and the combined totals for LLRTs demonstrate acceptable performance and support a conclusion that the structural integrity and leaktightness of the primary containment is adequately managed and will continue to be periodically monitored and managed effectively with the proposed changes.

Based on the above, the NRC staff finds that the licensee has addressed the NRC conditions to demonstrate acceptability of adopting NEI 94-01, Revision 3-A, and the limitations and conditions identified in the staff SE incorporated in NEI 94-01, Revision 2-A. The NRC also finds that the PRA used by the licensee is of sufficient technical adequacy to support the evaluation of changes to ILRT frequency. Therefore, the NRC staff concludes that the proposed changes to Braidwood and Byron TS 5.5.16 specified in Section 2.1 of this SE are acceptable and continue to meet 10 CFR 50.36(c)(5).

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Illinois State official was notified of the proposed issuance of the amendments on May 27, 2020. The State official 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 surveillance requirements. 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 Federal Register on December 17, 2019 (84 FR 68952), 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 Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors:

J. Dozier N. Chen R. Pettis G. Bedi Date of issuance: September 10, 2020

ML20149K698

• e-mail concurrence OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DEX/EMIB NRR/DSS/SCPB/BC NAME JWiebe*

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DATE 6/ 23 /2020 6/ 16 /2020 6/5/2020 3/10/2020 OFFICE NRR/DEX/ESEB NRR/DRA/ARCB/BC NRR/DSS/STSB/BC*

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JMcManus DATE 4/23/2020 5/5/2020 6/22/2020 8/10/2020 OFFICE NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME NSalgado*

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DATE 9/9/2020 9/10/2020