ML103270403

And Byron Station, Unit Nos. 1 and 2 - Issuance of Amendments Large Break Loss-of-Coolant Accident Analysis Using ASTRUM

ML103270403
Person / Time
Site:	Byron, Braidwood
Issue date:	12/21/2010
From:	Nicholas Difrancesco Plant Licensing Branch III
To:	Pacilio M Exelon Nuclear
DiFrancesco N, NRR/DORL/LPL3-2, 415-1115
Shared Package
ML103270423	List: ML103270403 ML103270435
References
TAC ME2941, TAC ME2942, TAC ME2943, TAC ME2944
Download: ML103270403 (27)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 21,2010 Mr. Michael J. Pacilio President and Chief Nuclear Officer 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 AMENDMENTS RE: LARGE BREAK LOSS-OF COOLANT ACCIDENT ANALYSIS USING THE AUTOMATED STATISTICAL TREATMENT OF UNCERTAINTY METHOD (TAC NOS. ME2941, ME2942, ME2943, AND ME2944)

Dear Mr. Pacilio:

The Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 164 to Facility Operating License No. NPF-72 and Amendment No. 164 to Facility Operating License No. NPF-77 for the Braidwood Station, Units 1 and 2 (Braidwood), and Amendment No. 170 to Facility Operating License No. NPF-37 and Amendment No. 170 to Facility Operating License No. NPF-66 for the Byron Station, Unit Nos. 1 and 2 (Byron), respectively.

The amendments are in response to your application dated December 16, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML093510099), as supplemented by letters dated April 26, and October 25,2010 (ADAMS Accession Nos.

ML101160431 and ML102980621, respectively). The amendments revise Technical Specifications Section 5.6.5, "Core Operating Limits Report (COLR)," to replace the existing reference for the large break loss-of-coolant accident (LOCA) analysis methodology with a reference to WCAP-16009-P-A, Revision 0, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM),"

January 2005.

M. Pacilio

- 2 A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Nicholas J. DiFrancesco, Project Manager Plant Licensing Branch 111-2 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. 164 to NPF-72

2. Amendment No. 164 to NPF-77

3. Amendment No. 170 to NPF-37

4. Amendment No. 170 to NPF-66

5. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-456 BRAIDWOOD STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 164 License No. NPF-72

1.

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

A.

The application for amendment by Exelon Generation Company, LLC (the licensee) dated December 16,2009, as supplemented by letters dated April 26, and October 25, 2010, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR) Chapter I; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-72 is hereby amended to read as follows:

- 2 (2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 164, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION IRA by E. Brown fori Robert D. Carlson, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Facility Operating License Date of Issuance: December 21,2010

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-457 BRAIDWOOD STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 164 License No. NPF-77

1.

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

A.

The application for amendment by Exelon Generation Company, LLC (the licensee) dated December 16, 2009, as supplemented by letters dated April 26, and October 25, 2010, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR) Chapter I; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-77 is hereby amended to read as follows:

- 2 (2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 164, and the Environmental Protection Plan contained in Appendix B, both of which are attached to License No. NPF-72, dated July 2, 1987, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION IRA by E. Brown forI Robert D. Carlson, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Facility Operating License Date of Issuance: December 21,2010

ATTACHMENT TO LICENSE AMENDMENT NOS. 164 AND 164 FACILITY OPERATING LICENSE NOS. NPF-72 AND NPF-77 DOCKET NOS. STN 50-456 AND STN 50-457 Replace the following pages of the 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 License NPF-72 License NPF-72 License Page 3 License Page 3 License NPF-77 License NPF-77 License Page 3 License Page 3 TSs TSs 5.6-4 5.6-4

- 3 (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.

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

(1)

Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels not in excess of 3586.6 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein and other items identified in Attachment 1 to this license. The items identified in to this license shall be completed as specified. is hereby incorporated into this license.

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 164, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Emergency Planning In the event that the NRC finds that the lack of progress in completion of the procedures in the Federal Emergency Management Agency's final rule, 44 CFR Part 350, is an indication that a major substantive problem exists in achieving or maintaining an adequate state of emergency preparedness, the provisions of 10 CFR Section 50.54(s)(2) will apply.

Amendment No. 164

-3 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 are 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 license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels not in excess of 3586.6 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein and other items identified in Attachment 1 to this license. The items identified in to this license shall be completed as specified. is hereby incorporated into this license.

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 164, and the Environmental Protection Plan contained in Appendix B, both of which are attached to License No. NPF-72, dated July 2, 1987, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and I

the Environmental Protection Plan.

I i (3)

Emergency Planning In the event that the NRC finds that the lack of progress in completion of the procedures in the Federal Emergency Management Agency's final rule, 44 CFR Part 350, is an indication that a major substantive problem exists in achieving or maintaining an adequate state of emergency preparedness, the provisions of 10 CFR Section 50.54(s){2) will apply_

Amendment No. 164

5.6 Reporting Requirements 5.6 Report"ing Requi rements 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

(continued)

5.

ComEd letter from D. Saccomando to the Office of Nuclear Reactor Regulation dated December 21, 1994, transmitting an attachment that documents applicable sections of WCAP-11992111993 and ComEd application of the UET methodology addressed in "Additional Information Regarding Application for Amendment to Facility Operating Licenses-Reactivity Control Systems."

6.

WCAP-16009-P-A, Revision 0, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," January 2005.

7.

WCAP-10079-P-A, "NOTRUMP, A Nodal Transient Small Break and General Network Code," August 1985.

8.

WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model using NOTRUMP Code," August 1985.

9.

WCAP-10216-P-A, Revision 1, "Relaxation of Constant Axial Offset Control - Fa Surveillance Technical Specification," February 1994.

10.

WCAP-8745-P-A, "Design Bases for the Thermal Overpower i1T and Thermal Overtemperature i1T Trip Functions,"

September 1986;

c.

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

d.

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

BRAIDWOOD - UNITS 1 &2 5.6 4

Amendment 164

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-454 BYRON STATION, UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 1 70 License No. NPF-37

1.

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

A.

The application for amendment by Exelon Generation Company, LLC (the licensee) dated December 16, 2009, as supplemented by letters dated April 26, and October 25, 2010, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR) Chapter I; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-37 is hereby amended to read as follows:

-2 (2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 170, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION IRA by E. Brown fori Robert D. Carlson, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Facility Operating License Date of Issuance: December 21,2010

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-455 BYRON STATION, UNIT NO.2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment NO.1 70 License No. NPF-66

1.

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

A.

The application for amendment by Exelon Generation Company, LLC (the licensee) dated December 16,2009, as supplemented by letters dated April 26, and October 25, 2010, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission'sJules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR) Chapter I; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-66 is hereby amended to read as follows:

- 2 (2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A (NUREG 1113), as revised through Amendment No. 170, and the Environmental Protection Plan contained in Appendix B, both of which are attached to License No. NPF-37, dated February 14,1985, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION IRA by E. Brown forI Robert D. Carlson, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Facility Operating License Date of Issuance: December 21, 2010

ATTACHMENT TO LICENSE AMENDMENT NOS. 170 AND 170 FACILITY OPERATING LICENSE NOS. NPF-37 AND NPF-66 DOCKET NOS. STN 50-454 AND STN 50-455 Replace the following pages of the 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 License NPF-37 License NPF-37 License Page 3 License Page 3 License NPF-66 License NPF-66 License Page 3 License Page 3 TSs TSs 5.6-4 5.6-4

-3 (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 license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels not in excess of 3586.6 megawatts thermal (100 percent power) in accordance with the conditions specified herein.

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 170, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Deleted.

(4)

Deleted.

(5)

Deleted.

(6)

The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the licensee's Fire Protection Report, and as approved in the SER dated February 1987 through Supplement No.8, subject to the following provision:

The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

Amendment No. 170

- 3 (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 are 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 license shall be deemed to contain and is subject to the conditions specified in the Commission's regulation 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 3586.6 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A (NUREG 1113), as revised through Amendment No. 170, and the Environmental Protection Plan contained in Appendix B, both of which are attached to License No. NPF-37, dated February 14, 1985, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Deleted.

(4)

Deleted.

(5)

Deleted.

Amendment No. 170

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

(continued)

5.

ComEd letter from D. Saccomando to the Office of Nuclear Reactor Regulation dated December 21, 1994, transmitting an attachment that documents applicable sections of WCAP-11992/11993 and ComEd application of the UET methodology addressed in "Additional Information Regarding Application for Amendment to Facility Operating Licenses-Reactivity Control Systems."

6.

WCAP-16009-P-A, Revision 0, "Realistic Large-Break LOCA Evaluation Methodology US-ing the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," January 2005.

7.

WCAP-10079 P-A, "NOTRUMP, A Nodal Transi ent Small Break and General Network Code," August 1985.

8.

WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation I~odel using NOTRUI"IP Code," August 1985.

9.

WCAP-10216 P-A, Revision 1, "Relaxation of Constant Axial Offset Control Fa Surveillance Technical Specification," February 1994.

10.

WCAP-8745-P-A, "Design Bases for the Thermal Overpower

~T and Thermal Overtemperature ~T Trip Functions,"

September 1986;

c.

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

d.

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

BYRON - UNITS 1 &2 5.6 - 4 Amendment 170

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 164 TO FACILITY OPERATING LICENSE NO. NPF-72, AMENDMENT NO. 164 TO FACILITY OPERATING LICENSE NO. NPF-77, AMENDMENT NO. 170 TO FACILITY OPERATING LICENSE NO. NPF-37, AND AMENDMENT NO. 170 TO 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 letter to the Nuclear Regulatory Commission (NRC, the Commission), dated December 16, 2009 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML093510099), as supplemented by letters dated April 26, and October 25,2010 (ADAMS Accession Nos. ML101160431 and ML102980621, respectively), Exelon Generation Company, LLC (EGC. the licensee), the licensee for Braidwood Station, Units 1 and 2 (Braidwood) and Byron Station, Unit Nos. 1 and 2 (Byron), submitted a request to implement the Westinghouse best-estimate loss-of-coolant accident (LOCA) analysis methodology known as the Automated Statistical Treatment of Uncertainty Method (ASTRUM), as documented in Westinghouse licensing topical report WCAP-16009-P-A, Revision 0, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," January 2005 (WCAP-16009-P-A). A non-proprietary copy is available at ADAMS Accession Nos. ML050910157 (cover letter), ML050910159 (Volume 1), and ML050910159 (Volume 2). The licensee requested to incorporate ASTRUM into its licensing basis large-break (LB) LOCA analysis, and to add a corresponding reference to WCAP-16009-P-A into Technical Specification (TS) 5.6.5, Core Operating Limits Report (COLR), Section b, which lists the COLR references.

The supplemental letters dated April 26 and October 25, 2010, provided additional information that clarified the application, did not exp&nd the scope of the application as originally noticed, and did not change NRC staff's original proposed no significant hazards consideration determination published in an individual notice in the Federal Register on February 23, 2010 (75 FR 8141).

-2

2.0 REGULATORY EVALUATION

ASTRUM is a NRC-approved method for demonstrating compliance with the acceptance criteria for emergency core cooling systems for light-water nuclear power reactors promulgated by Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 46 (10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors").

The regulation at 10 CFR 50.46(a)(1)(i) states, in part, that each pressurized light-water nuclear power reactor (PWR) must be provided with an emergency core cooling system (ECCS) that must be designed so that its calculated cooling performance following postulated LOCAs conforms to the criteria set forth in 10 CFR 50.46(b), that ECCS cooling performance must be calculated in accordance with an acceptable evaluation model, and that ECCS cooling performance must be calculated for a number of postulated LOCAs of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated LOCAs are calculated.

The acceptance criteria contained in 10 CFR 50.46(b), for which ASTRUM is used to demonstrate compliance, are 10 CFR 50.46(b)(1) - Peak cladding temperature - The calculated maximum fuel element cladding temperature shall not exceed 2200 degrees farenheit (OF).

10 CFR 50.46(b)(2) - Maximum cladding oxidation - The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

10 CFR 50.46(b)(3) - Maximum hydrogen generation - The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

Separate analyses and operator actions are credited to demonstrate compliance with the remaining 10 CFR 50.46(b) acceptance criteria, and these items are not affected by the implementation of the ASTRUM LBLOCA analysis.

Initial conditions assumed for the LOCA analyses are governed, in part, by TS Limiting Conditions for Operation (LCOs) and Surveillance Requirements (SRs). The requirements for TS are promulgated in 10 CFR 50.36, "Technical specifications." As specified by 10 CFR 50.36(c)(2), LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility. Pursuant to 10 CFR 50.36(c)(2)(ii)(B), an LCO is required for a process variable, design feature, or operating restriction that is an initial condition of a design-basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Also, 10 CFR 50.36(c)(3) defines an SR as a requirement relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

-3 The NRC staff considered the requirements of 10 CFR 50.36 to determine whether the initial conditions assumed in the LOCA analysis are acceptable.

3.0 TECHNICAL EVALUATION

Westinghouse obtained generic NRC approval of its original topical report describing the best-estimate LBLOCA methodology in 1996 for 3 and 4-loop PWRs. This method is known as the Code Qualification Document (CQD) methodology, documented in NRC-approved, Westinghouse Electric Corporation licensing topical report WCAP-12945-P-A, "Code Qualification Document for Best Estimate LOCA Analysis."

Westinghouse then completed a program to revise the statistical approach used to develop the Peak Cladding Temperature (PCT) and oxidation results at the 95th percentile. This method is based on the CQD methodology and follows the steps of the methodology in NUREG/CR-5249, "Quantifying Reactor Safety Margins - Application of Code Scaling, Applicability, and Uncertainty

[CSAU] Evaluation Methodology to a Large-Break, Loss-of-Coolant Accident (1989)." However, the uncertainty analysis (Element 3 in CSAU) is replaced by a technique based on order statistics. The ASTRUM methodology replaces the response surface technique with a statistical sampling method in which the uncertainty parameters are simultaneously sampled for each case. The approved ASTRUM evaluation model is documented in WCAP-16009-P-A.

3.1 Summary of Licensee's Analysis and Comparison to Current Licensing Basis EGC's current licensing basis LOCA analysis of record (AOR) for the Byron and Braidwood stations is based on the 1996 CQD methodology. The current licenSing basis PCTs for Byron and Braidwood Unit 1 are 2161 OF and 2168 of for Byron and Braidwood Unit 2.

The proposed ASTRUM analysis was performed for Byron and Braidwood operating at 3658.33 megawatts thermal (MWt) (i.e., at the current licensed thermal power level of 3586.6 MWt + 2 percent), and showed a limiting PCT of 2041 of for Byron and Braidwood Unit 2.

Exelon and its vendor, Westinghouse Electric Company, LLC, have interface processes which ensure that LOCA analysis input values conservatively bound current operating values.

The full results are tabulated below.

Parameter ASTRUM Results Braidwood &

Byron (Br/By)

Br/By Unit 2 Unit 1 10 CFR 50.46 Limits PCT 1913 OF 2041 of 2200 of Local Metal Oxidation 5.51%

8.27%

17%

Core-Wide Oxidation 0.25%

0.33%

1%

-4 3.2 NRC Staff Evaluation of Licensee's Analytic Results The NRC staff reviewed the information submitted by the licensee and concluded that the ASTRUM method is NRC-approved to analyze LBLOCAs at Westinghouse-designed, four loop plants such as Byron and Braidwood, and that the licensee's analysis demonstrates acceptable performance relative to the 10 CFR 50.46 acceptance criteria. In consideration of these items, the NRC staff finds the licensee's request to implement ASTRUM acceptable. A detailed discussion of the NRC staff's evaluation and conclusions follows.

3.2.1 Initial Conditions The NRC staff reviewed the licensee's assumed initial conditions to ensure that they are consistent with the current LOCA AOR, and thus, with the current Byron and Braidwood licensing bases. The NRC staff concluded that the initial conditions are consistent with the current licensing basis, with two notable exceptions.

EGC has, with this submittal, proposed to change the analyzed ranges of certain ECCS parameters, namely, the assumed accumulator pressure and the assumed safety injection temperature. The accumulator pressure range will be from 587 to 692 pounds per square inch absolute (psia), which is broadening of the previously-assumed range of 602 to 677 psia, which was listed in updated final safety analysis report Table 15.6-3b. Also, the assumed safety injection temperature range changes from 32 OF to 103 OF to 32 OF to 120 OF.

Accumulator Cover Pressure The NRC staff confirmed that the TS SRs, speCifically, SR 3.5.1.3 for Nitrogen Cover Pressure for the Safety Injection Accumulators will limit the actual values to a range that is more restrictive than those assumed in the accident analYSis. The following table compares TS allowable ranges for safety injection temperature (refueling water storage tank temperature) and accumulator pressure (safety injection accumulator nitrogen cover pressure) to the current LOCA AOR and the proposed ASTRUM analysis.1 Parameter TS Range AORRange ASTRUM Ranae Safety Injection Temp 35 °FsTsIs1 00 OF 32 °FsTs1S103 OF 32 °FSTs1S120 OF Accumulator Pressure 617psiasPaC;C:S;662psia 602psiasPaccs677psia 587psiasPaccs692psia Although it may seem that analyzing these parameters over a wider range than is permitted by the TS is a conservative analytic approach, the statistical nature of the ASTRUM method is such that multiple cases are sampled over the entire parametric range. Therefore, assuming the availability of a wider range of safety injection temperature than is permitted by the TS can result in the sampling of cases that are not reflective of operating conditions at the plant.

When the cases produce clearly conservative results, this approach is acceptable.

1 Although the TS requirement is specified in psi gage, the content of the license amendment request specified the accumulator pressure in psi absolute. The table displays all values in psi absolute, meaning that the staff added 15 psi to the TS values to account for atmospheric pressure.

- 5 However, assuming an increased accumulator cover pressure will result in a faster delivery of accumulator water to the reactor core. The NRC staff expects that this will artificially drive the core to a faster re-flood, thereby lowering the predicted PCT. For those cases that are sampled with a higher accumulator pressure than allowed by the plants TS, the NRC staff was concerned that the PCT results could be non-conservative.

Returning again to the discussion on the statistical nature of the ASTRUM method, if cases are sampled from the increased parametric ranges that result in a more optimistic prediction of accident performance, the overall effect could be an enhancement to the ASTRUM statistical base. Should this enhancement prove to be statistically significant, the level of assurance that ASTRUM has predicted a high-confidence upper limit on accident performance relative to the regulatory limits could be degraded.

WCAP-16009-P-A merely requires that the analyzed range of parameters for accumulator pressure and safety injection temperature bound TS requirements. Although WCAP-16009-P-A is approved with respect to demonstrating conformance to 10 CFR 50.46 requirements, the NRC staff must establish on a plant-specific basis that ASTRUM implementation meets 10 CFR 50.36 requirements. Therefore, on March 11, 2010 (ADAMS Accession No. ML100690463), the NRC staff requested that the licensee explain how the proposed, increased, analytic range of accumulator cover pressure remains in compliance with 10 CFR 50.36(c)(2)(ii)(B), since the analyzed range falls outside the range specified in the TS for operability.

In its April 26, 2010, response (ADAMS Accession No. ML101160431) to the NRC staff's request, EGC stated the following:

The approved WCAP 16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," evaluation model allows for the assumed accumulator pressure range to be partly outside the Technical Specification (TS) range, as discussed in WCAP-16009-P-A, Table 1-11, "Initial and Boundary Conditions Considered in Uncertainty Methodology." EGC elected to symmetrically widen the accumulator pressure range assumed in the analysis to facilitate evaluations of the significance of unanticipated events where accumulator pressure is outside the TS range, and to support potential future TS changes associated with accumulator pressure. The TS range is bounded by the accumulator pressure range assumed in the analysis.

The NRC staff believes that the 95/95 upper tolerance limit on PCT is based on the execution of an appropriate number of scenarios that apply to the existing plant configuration. Analytically exceeding TS limits by a reasonable amount provides assurance that the entire range of allowable parameters is covered in the plant analysis.

The licensee's response, however, did not assure the NRC staff that the increased parametric ranges would produce acceptably conservative results. Therefore, on September 10, 2010 (ADAMS Accession No. ML102240405), the NRC staff requested that the licensee provide additional information explaining how the increased parametric ranges assumed for accumulator pressure produced conservative results.

- 6 The licensee responded by letter dated October 25, 2010 (ADAMS Accession No. ML102980621).

The licensee stated that, while broadening the assumed parametric range of accumulator pressures would include cases where the accumulator pressure was high, and the resultant PGT thus possibly lower, the range was broadened symmetrically, which would mean that cases were also sampled with lower accumulator pressures, which would thus include possibly higher PCT results.

While the non-conservative PCTs would effectively lessen the lower tolerance limit, the conservative cases would produce a similarly increase in the upper tolerance limit. Therefore, because the broadening of assumed accumulator injection was symmetric, the result adds numerical dispersion in the base of predicted PCT s, and contributes to a more conservative overall result because of the increased upper tolerance limit. Because the licensee provided information to demonstrate that the overall effect of the symmetrically broadened accumulator initial pressures produces a conservative result, the NRC staff concludes that broadening the accumulator pressure range is acceptable.

Safety Injection Temperature The refueling water storage tank at Byron and Braidwood stations is required to remain between 35 OF and 100 of per TS SR 3.5.4.1. In the current licensing basis LOCA analysis, this range is bounded by an assumed range for safety injection temperature of 32 of to 103 of. For the ASTRUM analyses, the licensee increased the upper limit on this range to 120 of.

As stated in the October 25, 2010, supplemental letter, the lower bound on assumed safety injection temperature is at the freezing point of water, so the licensee did not symmetrically widen the assumed safety injection temperature range, but rather increased the upper limit by 1r F over the previous AOR, which is greater than the TS allowable value by 20 OF. Although this change is not symmetric, the NRC staff finds it acceptable because safety injection water would provide less capacity to cool the fuel, and thus this increased range would cause the PCT population to include cases with higher PCTs due to the higher safety injection temperature.

Conclusion - Initial Conditions LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility. Pursuant to 10 CFR 50.36(c)(2)(ii)(B), an LCO is required for a process variable, design feature, or operating restriction that is an initial condition of a design-basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Also, 10 CFR 50.36(c)(3) defines an SR as a requirement relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. The NRC staff finds that the licensee's selection of initial conditions for accumulator pressure and safety injection temperature are consistent with the requirements of 10 CFR 50.36(c)(2)(ii)(3) and 10 CFR 50.36(c)(3) because the licensee has demonstrated that, in both cases, the analytic results are conservative relative to the TS requirements.

-7 3.2.2 PCT Elevation The licensee included plots of key parameters from the limiting ASTRUM runs. The NRC staff observed that one of the results indicated that the PCT was located, from 200 seconds onward, at the 12 foot elevation for Byron and Braidwood Unit 1. On March 11, 2010, the NRC staff requested that the licensee review this result and provide additional information to confirm that this is an expected and acceptable result.

In its April 26, 2010, supplemental letter, the licensee explained that this limiting case included a top-peaked power shape. Additionally. the licensee explained that as a result of the quench front migrating from the bottom of the core to the top of the core during the re-flood phase, it is possible for the PCT elevation to remain very high in the core. The result provided represents a combination of these phenomena, which result in the peak clad temperature elevation being at the top of the core in later phases of the transient. The licensee also clarified that this was not the location of the limiting PCT, which occurred prior to 200 seconds into the transient, and at a lower elevation in the core. The NRC staff finds that the information that the licensee provided acceptably clarified the results depicted in the license amendment request, because it clarified that the figure in question was not identifying the actual, limiting PCT, but rather the location of the PCT at any given point in the transient.

3.3 Remaining Results The remaining results provided by the licensee appeared consistent with previous ASTRUM results for similar plants. The NRC staff did not identify any other, significant issues with the licensee's analytic results.

3.4 ASTRUM Implementation - NRC Staff Conclusion Based on the considerations discussed above, the NRC staff finds that the licensee's request to implement the ASTRUM LOCA evaluation model at the Byron and Braidwood Stations is acceptable. The NRC staff finds that the licensee's results meet the 10 CFR S0.46(b)(1-3) acceptance criteria, and that the analytic inputs are consistent with or conservative relative to the Byron and Braidwood current licensing basis, the TS, and, thus, 10 CFR SO.36 requirements.

4.0 STATE CONSULTATION

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

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

-8 (75 FR 8141, February 23,2010). 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) such activities will be conducted in compliance with the Commission's regulations; and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: B. Parks, NRR Date: December 21,2010

M. Pacilio

- 2 A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, IRA!

Nicholas J. DiFrancesco, Project Manager Plant Licensing Branch 111-2 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.164 to NPF-72

2. Amendment No.164 to NPF-77

3. Amendment No.170 to NPF-37

4. Amendment No.170 to NPF-66

5. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:

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