Text

Issuance of Amendments to Regarding Request to Delete Obsolete License Conditions and Make Administrative Changes to Technical Specifications (CAC Nos. MF9338-MF9341)
	ML17088A703
Person / Time
Site:	Byron, Braidwood
Issue date:	07/05/2017
From:	Joel Wiebe; Plant Licensing Branch III
To:	Bryan Hanson; Exelon Generation Co, Exelon Nuclear
	Wiebe J
References
	CAC MF9338, CAC MF9339, CAC MF9340, CAC MF9341
	Download: ML17088A703 (77)
	v • d • e

Mr. Bryan C. Hanson Senior Vice President UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 5, 2017 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 AMENDMENTS REGARDING REQUEST TO DELETE OBSOLETE LICENSE CONDITIONS AND MAKE ADMINISTRATIVE CHANGES TO TECHNICAL SPECIFICATIONS (CAC NOS. MF9338, MF9339, MF9340, AND MF9341)

Dear Mr. Hanson:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 193 to Renewed Facility Operating License No. NPF-72 and Amendment No. 193 to Renewed Facility Operating License No. NPF-77 for the, Braidwood Station, Units 1 and 2, respectively, and Amendment No. 198 to Renewed Facility Operating License No.

NPF-37 and Amendment No. 198 to Renewed Facility Operating License No. NPF-66 for the Byron Station, Unit Nos. 1 and 2, respectively. An Amendment No. 182 to Renewed Facility Operating License No. NPF-72 will not be issued to make the Amendment numbers for the Braidwood Station, Units 1 and 2, the same. The amendments are in response to your application dated February 23, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17055A631) as supplemented by your letter dated June 29, 2017 (ADAMS Accession No. ML17180A530).

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

Docket Nos. STN 50-456, STN 50-457, STN 50-454 and STN 50-455

Enclosures:

1. Amendment No.193 to NPF-72

2. Amendment No.193 to NPF-77

3. Amendment No. 1 98 to NPF-37

4. Amendment No.1 98 to NPF-66

5. Safety Evaluation cc w/encls: Distribution via Listserv s;J/Jj;J?

Joel S. Wiebe, Senior Project Manager Plant Licensing Branch 111 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY. LLC DOCKET NO. STN 50-456 BRAIDWOOD STATION. UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 193 Renewed License No. NPF-72

1.

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

A.

The application for amendment by Exelon Generation Company, LLC (the licensee) dated February 23, 2017, as supplemented by letter dated June 29, 2017, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, the Renewed Facility Operating License No. NPF-72 is amended to:

A.

In Appendix C, delete the following from Amendment Number 145 Additional Condition:

During operation in Cycles 15, 16, and 17, up to eight (8) AREVA NP Advanced Mark-BW(A) fuel assemblies containing fuel pellets incorporating homogeneous poisons may be placed in nonlimiting Unit 1 core locations provided the fuel cycle designs are developed such that the TS 2.1.1.3 Safety Limit equation for Westinghouse fuel is bounding. The design basis for the AREVA NP fuel rod centerline melt follows that given in BAW-10162P-A, "TAC03 - Fuel Pin Thermal Analysis Computer Code," October 1989, and BAW-10184P-A, "GDTACO - Urania Gadolinia Fuel Pin Thermal Analysis Code," February 1995.

B.

In Appendix C, delete the entire Amendment 146 Additional Condition C.

Change the Technical Specifications as indicated in the attachment to this license amendment, and paragraphs 2.C.(2) and 2.C.(7) of Renewed Facility Operating License No. NPF-72 are hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 193 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.

(7)

Additional Conditions The Additional Conditions contained in Appendix C, as revised through Amendment No. 193, are hereby incorporated into this renewed license.

The licensee shall operate the facility in accordance with the Additional Conditions.

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.

Attachment:

Changes to the Technical FOR THE NUCLEAR REGULA TORY COMMISSION Qdna2~ie~'-

Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Specifications and Renewed Facility Operating License Date of Issuance: July s, 2017

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-457 BRAIDWOOD STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 193 Renewed License No. NPF-77

1.

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

A The application for amendment by Exelon Generation Company, LLC (the licensee) dated February 23, 2017, as supplemented by letter dated June 29, 2017, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

The issuance of this amendment is in accordance with 1 O CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the Renewed Facility Operating License No. NPF-77 is amended to:

A In Appendix C, delete the entire Amendment 146 Additional Condition B.

Change the Technical Specifications as indicated in the attachment to this license amendment, and paragraphs 2.C.(2) and 2.C.(6) of Renewed Facility Operating License No. NPF-77 are hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 193 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.

(6)

Additional Conditions The Additional Conditions contained in Appendix C, as revised through Amendment No. 193, are hereby incorporated into this renewed license.

The licensee shall operate the facility in accordance with the Additional Conditions.

3.

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

Attachment:

Changes to the Technical FOR THE NUCLEAR REGULATORY COMMISSION

{)J 9 o/_

David J. Wrona, Chief Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Specifications and Renewed Facility Operating License Date of Issuance: July 5, 2017

ATTACHMENT TO LICENSE AMENDMENT NOS. 193 AND 193 RENEWED FACILITY OPERATING LICENSE NOS. NPF-72 AND NPF-77 DOCKET NOS. STN 50-456 AND STN 50-457 Replace the following pages of the Facility Operating Licenses and Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove License NPF-72 Page 3 Page 4 Appendix C, Page 1 Appendix C, Page 2 License NPF-77 Page 3 Page4 Appendix C, Page 2 TSs 2.0-1 3.0-4 3.0-6 3.1.7 -2 3.3.1-13 3.4.15-3 3.7.1 -1 3.7.8-1, Unit 1 3.7.8 - 2, Unit 1 3.7.8 - 3, Unit 1 3.7.8-1, Unit 2 3.7.8 - 2, Unit 2 3.7.8 - 3, Unit 2 3.7.8 - 4, Unit 2 3.7.10-3 3.8.4-2 3.8.5-2 3.8.8-2 4.0-1 4.0-2 5.5-21 License NPF-72 Page 3 Page4 Appendix C, Page 1 License NPF-77 Page 3 Page4 TSs 2.0-1 3.0-4 3.0-6 3.1.7 - 2 3.3.1-13 3.4.15 - 3 3.7.1 - 1 3.7.8 - 1 3.7.8 -2 3.7.8 - 3 3.7.10-3 3.8.4-2 3.8.5-2 3.8.8-2 4.0-1 4.0-2 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 Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

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

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 193 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. 193 (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 provision of 10 CFR Section 50.54(s)(2) will apply.

(4)

Deleted.

(5)

Deleted.

(6)

Deleted.

(7)

Additional Conditions The Additional Conditions contained in Appendix C, as revised through Amendment No. 193, are hereby incorporated into this renewed license.

The licensee shall operate the facility in accordance with the Additional Conditions.

(8)

Exelon Generation Company shall provide to the Director of the Office of Nuclear Reactor Regulation a copy of any application, at the time it is filed, to transfer (excluding grants of security interests or liens) from Exelon Generation Company to its direct or indirect parent, or to any other affiliated company, facilities for the production, transmission, or distribution of electric energy having a depreciated book value exceeding ten percent (10%) of Exelon Generation Company's consolidated net utility plant, as recorded on Exelon Generation Company's books of account.

(9)

Exelon Generation Company shall have decommissioning trust funds for Braidwood, Unit 1, in the following minimum amount, when Braidwood, Unit 1, is transferred to Exelon Generation Company:

Braidwood Unit 1

$154,273,345 Renewed License No. NPF-72 Amendment No. 193

APPENDIX C ADDITIONAL CONDITIONS FACILITY OPERATING LICENSE NO. NPF-72 The licensee shall comply with the following conditions on the schedules noted below:

Amendment Number 145 Additional Condition The safety limit equation specified in TS 2.1.1.3 regarding fuel centerline melt temperature (i.e., less than 5080 °F, decreasing by 58 °F per 10,000 MWD/MTU burn up as described in WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," April 1995) is valid for uranium oxide fuel without the presence of poisons mixed homogeneously into the fuel pellets. If fuel pellets incorporating homogeneous poisons are used, the topical report documenting the fuel centerline melt temperature basis must be reviewed and approved by the NRC and referenced in this license condition. TS 2.1.1.3 must be modified to also include the fuel centerline melt temperature limit for the fuel with homogeneous poison.

Implementation Date With imple-mentation of the amend-ment Amendment No. 193 (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 1 O CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

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

(1)

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

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 193 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. 193 (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 provision of 10 CFR Section 50.54(s)(2) will apply.

(4)

Deleted.

(5)

Deleted.

(6)

Additional Conditions The Additional Conditions contained in Appendix C, as revised through Amendment No. 193, are hereby incorporated into this renewed license.

The licensee shall operate the facility in accordance with the Additional Conditions.

(7)

Exelon Generation Company, LLC, shall provide to the Director of the Office of Nuclear Reactor Regulation, a copy of any application, at the time it is filed, to transfer (excluding grants of security interests or liens) from Exelon Generation Company, LLC, to its direct or indirect parent, or to any other affiliated company, facilities for the production, transmission, or distribution of electric energy having a depreciated book value exceeding ten percent (10%) of Exelon Generation Company's consolidated net utility plant, as recorded on Exelon Generation Company, LLC's books of account.

(8)

Exelon Generation Company, LLC, shall have decommissioning trust funds for Braidwood, Unit 2, in the following minimum amount, when Braidwood, Unit 2, is transferred to Exelon Generation Company, LLC:

Braidwood Unit 2

$154,448,967 (9)

The decommissioning trust agreement for Braidwood, Unit 2, at the time the transfer of the unit to Exelon Generation Company, LLC is effected and thereafter, is subject to the following:

Renewed License No. NPF-77 Amendment No. 193

SLs 2.0 2.0 SAFETY LIMITS CSLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded.

2.1.1.1 In MODE 1, the Departure from Nucleate Boiling Ratio (DNBR) shall be maintained 2 1.24 for the WRB-2 DNB correlation for a thimble cell, 2 1.25 for the WRB-2 DNB correlation for a typical cell and 2 1.19 for the ABB-NV DNB correlation for a thimble cell and a typical cell.

2.1.1.2 In MODE 2, the DNBR shall be maintained 2 1.17 for the WRB-2 DNB correlation, and 2 1.13 for the ABB-NV DNB correlation and 2 1.18 for the WLOP DNB correlation.

2.1.1.3 In MODES 1 and 2, the peak fuel centerline temperature shall be maintained< 5080°F decreasing by 58°F per 10,000 MWD/MTU burnup.

2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained

~ 2735 psig.

2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

BRAIDWOOD - UNITS 1 & 2 2.0 - 1 Amendment 193

LCD Applicability 3.0 3.0 LCD Applicability LCD 3.0.7 LCD 3.0.8 Exception LCOs allow specified Technical Specification (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged.

Compliance with Exception LCOs is optional.

When an Exception LCD is desired to be met but is not met, the ACTIONS of the Exception LCD shall be met.

When an Exception LCD is not desired to be met, entry into a MODE or other specified condition in the Applicability shall be made in accordance with the other applicable Specifications.

LCOs, including associated ACTIONs, shall apply to each unit individually, unless otherwise indicated. Whenever the LCD refers to a system or component that is shared by both units, the ACTIONs will apply to both units simultaneously.

BRAIDWOOD - UNITS 1 & 2 3.0 - 4 Amendment 193

SR Applicability 3.0 3.0 SR APPLICABILITY SR 3.0.3 SR 3.0.4 SR 3.0.5 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and the risk impact shall be managed.

If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 3.0.3.

When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4 This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

SRs shall apply to each unit individually, unless otherwise indicated.

BRAIDWOOD - UNITS 1 & 2 3.0 - 6 Amendment 193

Rod Position Indication 3.1. 7 ACTIONS (continued)

CONDITION REQUIRED ACTION C.

One demand position C.1.1 Verify by indicator per bank administrative means inoperable for one or all DRPis for the more banks.

affected bank(s) are OPERABLE.

AND C.1.2 Verify the most withdrawn rod and the least withdrawn rod of the affected bank( s) are

~ 12 steps apart.

OR C.2 Reduce THERMAL POWER to~ 50% RTP.

D.

Required Action and D.1 Be in MODE 3.

associated Completion Ti me not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.1. 7.1 Verify each DRPI agrees within 12 steps of the group demand position for the full indicated range of rod travel.

BRAIDWOOD - UNITS 1 & 2 3.1.7-2 COMPLETION TIME Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months FREQUENCY Prior to criticality after each removal of the reactor head Amendment 193

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.3.1.8 (continued)

SR 3. 3.1. 9

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

Verification of setpoint is not required.

Perform TADOT.

SR 3.3.1.10

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

This Surveillance shall include verification that the time constants are adjusted to the prescribed values.

Perform CHANNEL CALIBRATION.

SR 3.3.1.11

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

Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION.

BRAIDWOOD - UNITS 1 & 2 3.3.1-13 RTS Instrumentation 3.3.1 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program (continued)

Amendment 193

RCS Leakage Detection Instrumentation 3.4.15 ACTIONS (continued)

CONDITION REQUIRED ACTION D.

Required Action and D.1 Be in MODE 3.

associated Completion Time not met.

AND D.2 Be in MODE 5.

E.

All required monitors E.1 Enter LCO 3.0.3.

inoperable.

SURVEILLANCE REQUIREMENTS SR 3.4.15.1 SR 3.4.15.2 SR 3.4.15.3 SR 3.4.15.4 SURVEILLANCE Perform CHANNEL CHECK of the required containment atmosphere radioactivity monitor.

Perform COT of the required containment atmosphere radioactivity monitor.

Perform CHANNEL CALIBRATION of the required containment sump monitor.

Perform CHANNEL CALIBRATION of the required containment atmosphere radioactivity monitor.

BRAIDWOOD - UNITS 1 & 2 3.4.15 - 3 COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours Immediately FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Amendment 193

3.7 PLANT SYSTEMS 3.7.1 Main Steam Safety Valves (MSSVs)

LCO 3.7.1 Five MSSVs per steam generator shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS MSSVs 3.7.1

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

Separate Condition entry is allowed for each MSSV.

CONDITION A.

One or more steam A.1 generators with one or more MSSVs inoperable.

AND REQUIRED ACTION COMPLETION TIME Reduce THERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to less than or equal to the Maximum A 11 owa bl e % RTP specified in Table 3.7.1-1 for the number of OPERABLE MSSVs.

A.2

NOTE--------

BRAIDWOOD - UNITS 1 & 2 Only required in Mode 1.

Reduce the Power 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Range Neutron Flux -

High reactor trip setpoint to less than or equal to the Maximum Allowable

% RTP specified in Table 3.7.1-1 for the number of OPERABLE MSSVs.

(continued) I 3.7.1 - 1 Amendment 193

3.7 PLANT SYSTEMS 3.7.8 Essential Service Water (SX) System LCO 3.7.8 The following SX trains shall be OPERABLE:

a.

Two unit-specific SX trains; and SX System 3.7.8

b.

One opposite-unit SX train for unit-specific support.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION A. One unit-specific SX train inoperable.

BRAIDWOOD - UNITS 1 & 2 A.l REQUIRED ACTION

NOTES--------

1.

Enter applicable Conditions and Required Actions of LCO 3.8.1, "AC Sources-Operat i ng," for Emergency Diesel Generator made inoperable by SX.

2.

Enter applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops-MODE 4," for Residual Heat Removal loops made inoperable by sx.

COMPLETION TIME Restore unit-specific 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months SX train to OPERABLE status.

(continued) 3.7.8 - 1 Amendment 193

ACTIONS (continued)

CONDITION B.

Opposite-unit SX train B.1 inoperable.

C.

Required Action and C.1 associated Completion Time of Condition A or AND B not met.

C.2 BRAIDWOOD - UNITS 1 & 2 REQUIRED ACTION Restore opposite-unit SX train to OPERABLE status.

Be in MODE 3.

Be in MODE 5.

3.7.8 - 2 SX System 3.7.8 COMPLETION TIME 7 days 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours Amendment 193

SURVEILLANCE REQUIREMENTS SR 3.7.8.1 SR 3.7.8.2 SR 3.7.8.3 SR 3.7.8.4 SR 3.7.8.5 SU RV EI LLANCE

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

Isolation of SX flow to individual components does not render the SX System inoperable.

Verify each unit-specific SX manual, power operated, and automatic valve in the flow path servicing safety related equipment, that is not locked, sealed, or otherwise secured in position, is in the correct position.

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

Not required when opposite unit is in MODE 1, 2, 3, or 4.

Operate the opposite-unit SX pump for 2 15 minutes.

Cycle each opposite-unit SX crosstie valve that is not secured in the open position with power removed.

Verify each unit-specific SX automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

Verify each unit-specific SX pump starts automatically on an actual or simulated actuation signal.

BRAIDWOOD - UNITS 1 & 2 3.7.8 - 3 SX System 3.7.8 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Amendment 193

VC Filtration System 3.7.10 ACTIONS (continued)

CONDITION REQUIRED ACTION E.

Two VC Filtration E.1 Suspend movement of System trains irradiated fuel inoperable in MODE 5 assemblies.

or 6, or during movement of irradiated AND fuel assemblies.

E.2 Suspend positive OR reactivity additions.

One or more VC Filtration System trains inoperable due to an inoperable CRE boundary in MODE 5 or 6, or during movement of irradiated fuel assemblies.

F.

Two VC Filtration F.1 Enter LCO 3.0.3.

System trains inoperable in MODE 1, 2, 3, or 4 for reasons other than Condition B.

SURVEILLANCE REQUIREMENTS SR 3.7.10.1 SURVEILLANCE Operate each VC Filtration System train with:

a.

Flow through the makeup system filters for 2 15 continuous minutes with the heaters operating; and

b.

Flow through the recirculation charcoal adsorber for 2 15 minutes.

BRAIDWOOD - UNITS 1 & 2 3.7.10 - 3 COMPLETION TIME Immediately Immediately Immediately FREQUENCY In accordance with the Surveillance Frequency Control Program (continued)

Amendment 193

ACTIONS (continued)

CONDITION B.

One DC electrical power division crosstied to opposite-unit DC electrical power subsystem that has an inoperable battery charger, while opposite unit is in MODE 1, 2, 3, or 4.

c.

One DC electrical power division crosstied to opposite-unit DC electrical power subsystem with an inoperable source, while opposite unit is in MODE 5, 6, or defueled.

D.

One DC electrical power subsystem inoperable for reasons other than Condition A, B, or C.

E.

Required Action and Associated Completion Ti me not met.

BRAIDWOOD - UNITS 1 & 2 B.l C.l AND C.2 D.1 E.1 AND E.2 REQUIRED ACTION Open at least one crosstie breaker DC Sources-Operating 3.8.4 COMPLETION TI ME 204 hours0.00236 days 0.0567 hours 3.373016e-4 weeks 7.7622e-5 months between the crosstied divisions.

NOTE--------

Only required when opposite unit has an inoperable battery.

Verify opposite-unit Once per DC bus load 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

s: 200 amps.

Open at least one 7 days crosstie breaker between the crosstied divisions.

Restore DC electrical 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months power subsystem to OPERABLE status.

Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months 3.8.4 - 2 Amendment 193

ACTIONS CONDITION REQUIRED ACTION A.

(continued)

A.2.1 Suspend CORE ALTERATIONS.

AND A.2.2 Suspend movement of irradiated fuel assemblies.

AND A.2.3 Initiate action to suspend operations involving positive reactivity additions.

AND A.2.4 Initiate action to restore required DC electrical power subsystem to OPERABLE status.

AND A.2.5 Declare affected Low Temperature Overpressure Protection feature(s) inoperable.

BRAIDWOOD - UNITS 1 & 2 3.8.5 - 2 DC Sources-Shutdown 3.8.5 COMPLETION TIME Immediately Immediately Immediately Immediately Immediately (continued)

Amendment 193

ACTIONS Inverters-Shutdown 3.8.8 CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.2.1 Suspend CORE Immediately ALTERATIONS.

AND A.2.2 Suspend movement of Immediately irradiated fuel assemblies.

AND A.2.3 Initiate action to Immediately suspend operations involving positive reactivity additions.

AND A.2.4 Initiate action to Immediately restore required inverters to OPERABLE status.

AND A.2.5 Declare affected Low Immediately Temperature Overpressure Protection feature(s) inoperable.

SURVEILLANCE REQUIREMENTS SR 3.8.8.1 SU RV EI LLANCE FREQUENCY Verify correct inverter voltage and breaker In accordance alignment to required AC instrument buses.

with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 & 2 3.8.8 - 2 Amendment 193

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site 4.1.1 Site Location The site is located in Reed Township, approximately 20 mi (32 km) south-southwest of the city of Joliet in northern Illinois.

4.1.2 Exclusion Area Boundary (EAB)

The EAB shall not be less than 1591 ft (485 meters) from the outer containment wall.

4.1.3 Low Population Zone (LPZ)

The LPZ shall be a 1.125 mi (1811 meter) radius measured from the midpoint between the two reactors.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 193 fuel assemblies.

Each assembly shall consist of a matrix of Zircaloy, ZIRLO, or Optimized ZIRLO' clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material.

Limited substitutions of zirconium alloy or stainless steel filler rods or vacancies for fuel rods, in accordance with approved applications of fuel rod configurations, may be used.

Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

4.2.2 Control Rod Assemblies The reactor core shall contain 53 control rod assemblies.

The control material shall be silver indium cadmium, hafnium, or a mixture of both types.

BRAIDWOOD - UNITS 1 & 2 4.0 - 1 Amendment 193

4.0 DESIGN FEATURES 4.3 Fuel Storage 4.3.1 Criticality Design Features 4.0 The spent fuel storage racks are designed and shall be maintained, as applicable, with:

a.

Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent;

b.

For Holtec spent fuel pool storage racks,

~tt ~ 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Holtec International Report HI-982094, "Criticality Analysis for Byron/Braidwood Rack Installation Project," Project No. 80944, 1998;

c.

For Holtec spent fuel pool storage racks, a nominal 10.888 inch north-south and 10.574 inch east-west center to center distance between fuel assemblies placed in Region 1 racks; and

d.

For Holtec spent fuel pool storage racks, a nominal 8.97 inch center to center distance between fuel assemblies placed in Region 2 racks.

4.3.2 Drainage The spent fuel pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 410 ft, 0 inches.

4.3.3 Capacity The spent fuel pool is designed and shall be maintained with a storage capacity limited to no more than 2984 fuel assemblies.

BRAIDWOOD - UNITS 1 & 2 4.0 - 2 Amendment 193

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 5.5.16 Safety Function Determination Program CSFDP)

(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.

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(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, September 1995 and NEI 94-01, Revision 0.

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 s 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 Band C tests and< 0.75 La for Type A tests; and BRAIDWOOD - UNITS 1 & 2 5.5 - 21 Amendment 193

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-454 BYRON STATION. UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERA TING LICENSE Amendment No. 198 Renewed 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 February 23, 2017, as supplemented by letter dated June 29, 2017, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, the Renewed Facility Operating License No. NPF-37 is amended to:

A.

In Appendix C, delete the entire Amendment 151 Additional Condition B.

Change the Technical Specifications as indicated in the attachment to this license amendment, and paragraphs 2.C.(2) and 2.C.(17) of Renewed Facility Operating License No. NPF-37 are hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 198 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.

(17)

Additional Conditions The Additional Conditions contained in Appendix C, as revised through Amendment No. 198, are hereby incorporated into this renewed license.

The licensee shall operate the facility in accordance with the Additional Conditions.

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.

Attachment:

Changes to the Technical FOR THE NUCLEAR REGULA TORY COMMISSION

~'.\\ UJ9 David J. Wrona, Chief Plant Licensing Branch 111 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Specifications and Renewed Facility Operating License Dateoflssuance:July 5, 2017

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY. LLC DOCKET NO. STN 50-455 BYRON STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 198 Renewed 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 February 23, 2017, as supplemented by letter dated June 29, 2017, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, the Renewed Facility Operating License No. NPF-66 is amended to:

A.

In Appendix C, delete the entire Amendment 151 Additional Condition B.

Change the Technical Specifications as indicated in the attachment to this license amendment, and paragraphs 2.C.(2) and 2.C.(6) of Renewed Facility Operating License No. NPF-66 are hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A (NUREG-1113), as revised through Amendment No. 198 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.

(6)

Additional Conditions The Additional Conditions contained in Appendix C, as revised through Amendment No. 198, are hereby incorporated into this renewed license.

The licensee shall operate the facility in accordance with the Additional Conditions.

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.

Attachment:

Changes to the Technical FOR THE NUCLEAR REGULATORY COMMISSION cl_~/ <J David J. Wrona, Chief Plant Licensing Branch 111 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Specifications and Renewed Facility Operating License Date of Issuance: July 5, 2017

ATTACHMENT TO LICENSE AMENDMENT NOS. 198 AND 198 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 License and Appendix A Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove License NPF-37 Page 3 Page4 Appendix C, Page 2 License NPF-66 Page 3 Page4 Appendix C, Page 2 TSs 2.0-1 3.0-4 3.0-6 3.1.7-2 3.3.1 -14 3.4.15-3 3.7.1 - 1 3.7.1-2 3.7.8-1 3.7.8 -2 3.7.9-2 3.7.10-3 3.7.10-4 3.8.4-2 3.8.5-2 4.0-2 4.0-3 5.5-21 Insert License NPF-37 Page 3 Page4 License NPF-66 Page 3 Page4 TSs 2.0-1 3.0-4 3.0-6 3.1.7-2 3.3.1 -14 3.4.15-3 3.7.1 - 1 3.7.1-2 3.7.8-1 3.7.8-2 3.7.9-2 3.7.10-3 3.7.10-4 3.8.4-2 3.8.5-2 4.0-2 5.5-21 (2)

Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Updated Final Safety Analysis Report, as supplemented and amended; (3)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

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

(1)

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

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 198 and the Environmental Protection Plan contained in Appendix 8, 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. 198 (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:

(7)

(8)

(9)

(10)

(11)

(12)

(13)

(14)

(15)

(16)

(17)

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.

Deleted Deleted.

Deleted.

AddWonalCondWons The Additional Conditions contained in Appendix C, as revised through Amendment No. 198, are hereby incorporated into this renewed license.

The licensee shall operate the facility in accordance with the Additional Conditions.

Renewed License No. NPF-37 Amendment No. 198 (2)

Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Updated Final Safety Analysis Report, as supplemented and amended; (3)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

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

(1)

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

(2)

Technical Specifications The Technical Specifications contained in Appendix A (NUREG-1113), as revised through Amendment No. 198 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. 198 (3)

Deleted.

(4)

Deleted.

(5)

Deleted.

(6)

Additional Conditions The Additional Conditions contained in Appendix C, as revised through Amendment No. 198, are hereby incorporated into this renewed license.

The licensee shall operate the facility in accordance with the Additional Conditions.

(7)

Exelon Generation Company, LLC shall provide the Director of the Office of Nuclear Reactor Regulation, a copy of any application, at the time it is filed, to transfer (excluding grants of security interests or liens) from Exelon Generation Company, LLC to its direct or indirect parent, or to any other affiliated company, facilities for the production, transmission, or distribution of electric energy having a depreciated book value exceeding ten percent (10%) of Exelon Generation Company, LLC's consolidated net utility plant, as recorded on Exelon Generation Company, LLC's books of account.

(8)

Exelon Generation Company, LLC shall have decommissioning trust funds for Byron, Unit 2, in the following minimum amount, when Byron, Unit 2, is transferred to Exelon Generation Company, LLC:

Byron Unit 2

$156,560,489 (9)

The decommissioning trust agreement for Byron, Unit 2, at the time the transfer of the unit to Exelon Generation Company, LLC is effected and thereafter, is subject to the following:

(a)

The decommissioning trust agreement must be in a form acceptable to the NRC.

(b)

With respect to the decommissioning trust fund, investments in the securities or other obligations of Exelon Corporation or affiliates thereof, or their successors or assigns are prohibited. Except for investments tied to market indexes or other non-nuclear sector mutual funds, investments in any entity owning one or more nuclear power plants are prohibited.

Renewed License No. NPF-66 Amendment No. 198

SLs 2.0 2.0 SAFETY LIMITS CSLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded.

2.1.1.1 In MODE 1, the Departure from Nucleate Boiling Ratio CDNBR) shall be maintained 2 1.24 for the WRB-2 DNB correlation for a thimble cell, 2 1.25 for the WRB-2 DNB correlation for a typical cell and 2 1.19 for the ABB-NV DNB correlation for a thimble cell and a typical cell.

2.1.1.2 In MODE 2, the DNBR shall be maintained 2 1.17 for the WRB-2 DNB correlation, and 2 1.13 for the ABB-NV DNB correlation and 2 1.18 for the WLOP DNB correlation.

2.1.1.3 In MODES 1 and 2, the peak fuel centerline temperature shall be maintained< 5080°F, decreasing by 58°F per 10,000 MWD/MTU burnup.

2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained

~ 2735 psig.

2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

BYRON - UNITS 1 & 2 2.0 - 1 Amendment 198

LCO Applicability 3.0 3.0 LCO Applicability LCO 3.0.7 LCO 3.0.8 Exception LCOs allow specified Technical Specification CTS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged.

Compliance with Exception LCOs is optional.

When an Exception LCO is desired to be met but is not met, the ACTIONS of the Exception LCO shall be met.

When an Exception LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall be made in accordance with the other applicable Specifications.

LCOs, including associated ACTIONs, shall apply to each unit individually, unless otherwise indicated. Whenever the LCO refers to a system or component that is shared by both units, the ACTIONs will apply to both units simultaneously.

BYRON - UNITS 1 & 2 3.0 - 4 Amendment 198

SR Applicability 3.0 3.0 SR APPLICABILITY SR 3.0.3 SR 3.0.4 SR 3.0.5 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and the risk impact shall be managed.

If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 3.0.3.

When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4.

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

SRs shall apply to each unit individually, unless otherwise indicated.

BYRON - UNITS 1 & 2 3.0 - 6 Amendment 198

Rod Position Indication 3.1. 7 ACTIONS (continued)

CONDITION REQUIRED ACTION

c.

One demand position C.1.1 Verify by indicator per bank administrative means inoperable for one or all DRPis for the more banks.

affected bank(s) are OPERABLE.

AND C.1.2 Verify the most withdrawn rod and the least withdrawn rod of the affected bank(s) are

~ 12 steps apart.

OR C.2 Reduce THERMAL POWER to~ 50% RTP.

D.

Required Action and D.1 Be in MODE 3.

associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.1. 7.1 Verify each DRPI agrees within 12 steps of the group demand position for the full indicated range of rod travel.

BYRON - UNITS 1 & 2 3.1.7-2 COMPLETION TIME Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months FREQUENCY Prior to criticality after each removal of the reactor head Amendment 198

SURVEILLANCE REQUIREMENTS SU RV EI LLANCE SR 3.3.1.8 (continued)

SR 3.3.1.9

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

Verification of setpoint is not required.

Perform TADOT.

SR 3.3.1.10

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

This Surveillance shall include verification that the time constants are adjusted to the prescribed values.

Perform CHANNEL CALIBRATION.

SR 3.3.1.11

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

Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION.

BYRON - UNITS 1 & 2 3.3.1 -

14 RTS Instrumentation 3.3.1 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program (continued)

Amendment 198

RCS Leakage Detection Instrumentation 3.4.15 ACTIONS (continued)

CONDITION REQUIRED ACTION

0.

Required Action and 0.1 Be in MOOE 3.

associated Completion Ti me not met.

ANO 0.2 Be in MOOE 5.

E.

All required monitors E.1 Enter LCD 3.0.3.

inoperable.

SURVEILLANCE REQUIREMENTS SR 3.4.15.1 SR 3.4.15.2 SR 3.4.15.3 SR 3.4.15.4 SURVEILLANCE Perform CHANNEL CHECK of the required containment atmosphere radioactivity monitor.

Perform COT of the required containment atmosphere radioactivity monitor.

Perform CHANNEL CALIBRATION of the required containment sump monitor.

Perform CHANNEL CALIBRATION of the required containment atmosphere radioactivity monitor.

BYRON - UNITS 1 & 2 3.4.15-3 COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours Immediately FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Amendment 198

3.7 PLANT SYSTEMS 3.7.1 Main Steam Safety Valves (MSSVs)

LCD 3.7.1 Five MSSVs per steam generator shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS MSSVs 3.7.1

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

Separate Condition entry is allowed for each MSSV.

CONDITION A.

One or more steam A.1 generators with one or more MSSVs inoperable.

AND REQUIRED ACTION COMPLETION TIME Reduce THERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to less than or equal to the Maximum A 11 owa bl e % RTP specified in Table 3.7.1-1 for the number of OPERABLE MSSVs.

A.2

NOTE--------

BYRON - UNITS 1 & 2 Only required in Mode 1.

Reduce the Power 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Range Neutron Flux-High reactor trip setpoint to less than or equal to the Maximum Allowable

% RTP specified in Table 3.7.1-1 for the number of OPERABLE MSSVs.

(continued) I 3.7.1 - 1 Amendment 198

ACTIONS (continued)

CONDITION B.

Required Action and B.1 associated Completion Time not met.

AND OR B.2 One or more steam generators with ~ 4 MSSVs inoperable.

SURVEILLANCE REQUIREMENTS REQUIRED ACTION Be in MODE 3.

Be in MODE 4.

SU RV EI LLANCE SR 3. 7.1.1

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

Only required to be performed in MODES 1 and 2.

Verify each required MSSV lift setpoint per Table 3.7.1-2 in accordance with the INSERVICE TESTING PROGRAM.

Following testing, lift setting shall be within+/- 1%.

BYRON - UNITS 1 & 2 3.7.1 - 2 MSSVs 3.7.1 COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 12 hours FREQUENCY In accordance with the INSERVICE TESTING PROGRAM Amendment 198

3.7 PLANT SYSTEMS 3.7.8 Essential Service Water (SX) System LCO 3.7.8 The following SX trains shall be OPERABLE:

a.

Two unit-specific SX trains; and SX System 3.7.8

b.

One opposite-unit SX train for unit-specific support.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION A.

One unit-specific SX train inoperable.

BYRON - UNITS 1 & 2 A.1 REQUIRED ACTION

NOTES--------

1.

Enter applicable Conditions and Required Actions of LCO 3.8.1, "AC Sources-Operat i ng," for Emergency Diesel Generator made inoperable by SX.

2.

Enter applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops-MODE 4," for Residual Heat Removal loops made inoperable by sx.

COMPLETION TIME Restore unit-specific 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months SX train to OPERABLE status.

(continued) 3.7.8 - 1 Amendment 198

ACTIONS (continued)

CONDITION B.

Opposite-unit SX train B.1 inoperable.

c.

Required Action and C.l associated Completion Time of Condition A or AND B not met.

C.2 BYRON - UNITS 1 & 2 REQUIRED ACTION Restore opposite-unit SX train to OPERABLE status.

Be in MODE 3.

Be in MODE 5.

3.7.8 - 2 SX System 3.7.8 COMPLETION TIME 7 days 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours Amendment 198

ACTIONS (continued)

CONDITION

c. Outside air wet bulb C.l temperature> 76°F.

AND Any electrical division not capable AND of providing power to at least one OPERABLE C.2 SXCT fan.

D.

SX pump discharge D.1 water temperature

> 96°F.

AND D.2 E.

One or more basin E.1 level(s) < 60%.

BYRON - UNITS 1 & 2 REQUIRED ACTION Verify OPERABLE SXCT fans are capable of being powered by an OPERABLE emergency power source.

Restore SXCT fan configuration such that each electrical division is capable of providing power to at least one OPERABLE SXCT fan.

Be in MODE 3.

Be in MODE 5.

Restore both basin levels to~ 60%.

3.7.9 - 2 UHS 3.7.9 COMPLETION TIME 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 72 hours 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (continued)

Amendment 198

VC Filtration System 3.7.10 ACTIONS (continued)

CONDITION REQUIRED ACTION E.

Two VC Filtration E.1 Suspend movement of System trains irradiated fuel inoperable in MODE 5 assemblies.

or 6, or during movement of AND irradiated fuel assemblies.

E.2 Suspend positive reactivity additions.

OR One or more VC Filtration System trains inoperable due to an inoperable CRE boundary in MODE 5 or 6, or during movement of irradiated fuel assemblies.

F.

Two VC Filtration F.1 Enter LCO 3.0.3.

System trains inoperable in MODE 1, 2, 3, or 4 for reasons other than Condition B.

SURVEILLANCE REQUIREMENTS SR 3.7.10.1 SURVEILLANCE Operate each VC Filtration System train with:

a.

Flow through the makeup system filters for ~ 15 continuous minutes with the heaters operating; and

b.

Flow through the recirculation charcoal adsorber for ~ 15 minutes.

BYRON - UNITS 1 & 2 3.7.10 - 3 COMPLETION TIME Immediately Immediately Immediately FREQUENCY In accordance with the Surveillance Frequency Control Program (continued)

Amendment 198

VC Filtration System

3. 7.10 SURVEILLANCE REQUIREMENTS (continued)

SR 3.7.10.2 SR 3.7.10.3 SR 3.7.10.4 SURVEILLANCE Perform required VC Filtration System filter testing in accordance with the Ventilation Filter Testing Program (VFTP).

Verify each VC Filtration System train actuates on an actual or simulated actuation signal.

Perform required CRE unfiltered air inleakage testing in accordance with the Control Room Envelope Habitability Program.

BYRON - UNITS 1 & 2 3.7.10 - 4 FREQUENCY In accordance with the VFTP In accordance with the Surveillance Frequency Control Program In accordance with the Control Room Envelope Habitability Program Amendment 198

ACTIONS (continued)

CONDITION B.

One DC electrical power division crosstied to opposite-unit DC electrical power subsystem that has an inoperable battery charger, while opposite unit is in MODE 1, 2, 3, or 4.

C.

One DC electrical power division crosstied to opposite-unit DC electrical power subsystem with an inoperable source, while opposite unit is in MODE 5, 6, or defueled.

D.

One DC electrical power subsystem inoperable for reasons other than Condition A, B, or C.

E.

Required Action and Associated Completion Time not met.

BYRON - UNITS 1 & 2 B.l C.l AND C.2 D.l E.l AND E.2 REQUIRED ACTION Open at least one crosstie breaker DC Sources-Operating 3.8.4 COMPLETION TIME 204 hours0.00236 days 0.0567 hours 3.373016e-4 weeks 7.7622e-5 months between the crosstied divisions.

NOTE--------

Only required when opposite unit has an inoperable battery.

Verify opposite-unit Once per DC bus load 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

200 amps.

Open at least one crosstie breaker 7 days between the crosstied divisions.

Restore DC electrical 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months power subsystem to OPERABLE status.

Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months 3.8.4 - 2 Amendment 198

ACTIONS CONDITION REQUIRED ACTION A.

(continued)

A.2.1 Suspend CORE ALTERATIONS.

AND A.2.2 Suspend movement of irradiated fuel assemblies.

AND A.2.3 Initiate action to suspend operations involving positive reactivity additions.

AND A.2.4 Initiate action to restore required DC electrical power subsystem to OPERABLE status.

AND A.2.5 Declare affected Low Temperature Overpressure Protection feature(s) inoperable.

BYRON - UNITS 1 & 2 3.8.5 - 2 DC Sources-Shutdown 3.8.5 COMPLETION TIME Immediately Immediately Immediately Immediately Immediately (continued)

Amendment 198

4.0 DESIGN FEATURES 4.3 Fuel Storage 4.3.l Criticality Design Features 4.0 The spent fuel storage racks are designed and shall be maintained, as applicable, with:

a.

Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent;

b.

A ken~ 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Holtec International Report HI-982094, "Criticality Analysis for Byron/Braidwood Rack Installation Project," Project No.

80944' 1998;

c.

A nominal 10.888 inch north-south and 10.574 inch east-west center to center distance between fuel assemblies placed in Region 1 racks; and

d.

A nominal 8.97 inch center to center distance between fuel assemblies placed in Region 2 racks.

4.3.2 Drainage The spent fuel pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 410 ft, 0 inches.

4.3.3 Capacity The spent fuel pool is designed and shall be maintained with a storage capacity limited to no more than 2984 fuel assemblies.

BYRON - UNITS 1 & 2 4.0 - 2 Amendment 198

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 5.5.16 Safety Function Determination Program CSFDP)

(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.

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(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, September 1995 and NEI 94-01, Revision 0.

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 Band C tests and< 0.75 La for Type A tests; and BYRON - UNITS 1 & 2 5.5 - 21 Amendment 198

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 193 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-72, AMENDMENT NO. 193 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-77.

AMENDMENT NO. 198 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-37.

AND AMENDMENT NO. 198 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 letter to the U.S. Nuclear Regulatory Commission (NRC, the Commission) dated February 23, 2017 (Agencywide Documents Access and Management System Accession No. ML17055A631 ), as supplemented by letter dated June 29, 2017 (ADAMS Accession No. ML17180A530), Exelon Generation Company, LLC (the licensee) requested changes to the facility operating licenses and technical specification (TSs) for the Braidwood Station (Braidwood), Units 1 and 2, and Byron Station (Byron), Unit Nos. 1 and 2.

The amendment removes time, cycle, or modification-related items from the operating licenses (OLs) and technical specifications (TS). Additionally, the amendment makes editorial and formatting changes. The licensee's basis for the amendment is that the time, cycle, or modification-related items have been implemented or superseded, are no longer applicable, and no longer need to be maintained in their associated OLs or TS.

A proposed determination of no significant hazards was published in the Federal Register on April 11, 2017. The supplement, dated June 29, 2017, contained clarifying information and did not change the scope of the proposed action or affect the NRC staff's initial proposed finding of no significant hazards consideration.

2.0 REGULATORY EVALUATION

Section 182a of the Atomic Energy Act of 1954, as amended, requires applicants for nuclear power plant OLs to include TSs as part of the license. The NRC's regulatory requirements related to the content of the TSs are set forth in Title 1 O of the Code of Federal Regulations (10 CFR) Section 50.36, Technical specifications." This regulation requires that TSs include:

(1) safety limits, limiting safety system settings and limiting control settings, (2) limiting conditions for operation, (3) surveillance requirements (SRs), (4) design features, and (5) administrative controls.

Section 187 of the Atomic Energy Act, "Modification of License," states that the "terms and conditions of all licensees shall be subject to amendment, revision, or modification, by reason of amendments of this Act, or by reason of rules and regulations issued in accordance with the terms of this Act." This provision authorizes NRC to amend licenses.

Section 50.50 of 10 CFR states, in part, that the Commission will issue a license in such form and containing such conditions and limitations, including TSs, as it deems appropriate and necessary.

3.0 TECHNICAL EVALUATION

3.1 License Conditions Braidwood, Unit 1, Appendix C, Additional Conditions, Facility Operating License No. NPF-72 Current License Condition:

Amendment Number 145 Additional Condition The safety limit equation specified in TS 2.1.1.3 regarding fuel centerline melt temperature (i.e., less than 5080 °F, decreasing by 58 °F per 10,000 MWD/MTU (megawatt days per metric ton of uranium] burnup as described in WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report, 11 April 1995) is valid for uranium oxide fuel without the presence of poisons mixed homogeneously into the fuel pellets. If fuel pellets incorporating homogeneous poisons are used, the topical report documenting the fuel centerline melt temperature basis must be reviewed and approved by the NRC and referenced in this license condition. TS 2.1.1.3 must be modified to also include the fuel centerline melt temperature limit for the fuel with homogeneous poison.

During operation in Cycles 15, 16, and 17, up to eight (8)

AREVA NP Advanced Mark-BW(A) fuel assemblies containing fuel pellets incorporating homogeneous poisons may be placed in nonlimiting Unit 1 core locations provided the fuel cycle designs are developed such that the TS 2.1.1.3 Safety Limit equation for Westinghouse fuel is bounding. The design basis for the AREVA NP fuel rod centerline melt follows that given in BAW-10162P-A, "TAC03 - Fuel Pin Thermal Analysis Computer Code,"

October 1989, and BAW-10184P-A, "GDTACO - Urania Gadolinia Fuel Pin Thermal Analysis Code, 11 February 1995.

Implementation Date With implementation of the amendment Amended License Condition:

Amendment Number Additional Condition Implementation Date 145 The safety limit equation specified in TS 2.1.1.3 regarding fuel centerline melt temperature (i.e., less than 5080 °F, decreasing by 58 °F per 10,000 MWD/MTU burnup as described in WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," April 1995), is valid for uranium oxide fuel without the presence of poisons mixed homogeneously into the fuel pellets. If fuel pellets incorporating homogeneous poisons are used, the topical report documenting the fuel centerline melt temperature basis must be reviewed and approved by the NRC and referenced in this license condition. TS 2.1.1.3 must be modified to also include the fuel centerline melt temperature limit for the fuel with homogeneous poison.

With implementation of the amendment The following text is deleted from the amended license condition:

During operation in Cycles 15, 16, and 17, up to eight (8) AREVA NP Advanced Mark-BW(A) fuel assemblies containing fuel pellets incorporating homogeneous poisons may be placed in nonlimiting Unit 1 core locations provided the fuel cycle designs are developed such that the TS 2.1.1.3 Safety Limit equation for Westinghouse fuel is bounding. The design basis for the AREVA NP fuel rod centerline melt follows that given in BAW-10162P-A, "TAC03 - Fuel Pin Thermal Analysis Computer Code," October 1989, and BAW-10184P-A, "DTACO - Urania Gadolinia Fuel Pin Thermal Analysis Code," February 1995.

The licensee states in its February 23, 2017, letter that operating cycles 15, 16, and 17 are complete for Braidwood, Unit 1. The NRC staff finds that since the window of operating cycles for which the license condition applied has expired, the identified text is no longer necessary.

The NRC staff, therefore, determines that 10 CFR 50.50 is satisfied with the identified text removed.

Current License Condition:

Amendment Number 146 Additional Condition Upon implementation of Amendment No. 146 adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 3.7.10.4, in accordance with TS 5.5.18.c.(i), the assessment of CRE habitability as required by Specification 5.5.18.c.(ii), and the measurement of CRE pressure as required by Specification 5.5.18.d, shall be considered met. Following implementation:

Implementation Date With implementation of the amendment (a) The first performance of SR 3.7.10.4, in accordance with Specification 5.5.18.c.(i), shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from November 7, 2004, the date of the most recent successful tracer gas test, as stated in the February 7, 2005 letter response to Generic Letter 2003-01, or within the next 18 months if the time period since the most recent successful tracer gas test is greater than 6 years.

(b) The first performance of the periodic assessment of CRE habitability, Specification 5.5.18.c.(ii), shall be within 3 years, plus the 9-month allowance of SR 3.0.2, as measured from November 7, 2004, the date of the most recent successful tracer gas test as stated in the February 7, 2005 letter response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

(c) The first performance of the periodic measurement of CRE pressure, Specification 5.5.18.d, shall be within 18 months, plus the 138 days allowed by SR 3.0.2, as measured from the date of the most recent successful pressure measurement test, or within 138 days if not performed previously.

Amended License Condition:

This license condition is deleted in its entirety.

The licensee states in its February 23, 2017, letter that the first performances of the CRE habitability surveillance, assessment, and measurement as required by the license condition items (a), (b), and (c) respectively are complete. The NRC staff finds that after the completion of the first performances of (a), (b), and (c) the license condition is no longer necessary.

Therefore the NRC staff determines that 1 O CFR 50.50 is met with this license condition deleted.

Braidwood, Unit 2, Appendix C, Additional Conditions, Facility Operating License No. NPF-77 Current License Condition:

Amendment Number 146 Additional Condition Upon implementation of Amendment No. 146 adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 3.7.10.4, in accordance with TS 5.5.18.c.(i), the Implementation Date With implementation of the amendment assessment of CRE habitability as required by Specification 5.5.18.c.(ii), and the measurement of CRE pressure as required by Specification 5.5.18.d, shall be considered met. Following implementation:

(a) The first performance of SR 3.7.10.4, in accordance with Specification 5.5.18.c.(i), shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from November 7, 2004, the date of the most recent successful tracer gas test, as stated in the February 7, 2005 letter response to Generic Letter 2003-01, or within the next 18 months if the time period since the most recent successful tracer gas test is greater than 6 years.

(b) The first performance of the periodic assessment of CRE habitability, Specification 5.5.18.c.(ii), shall be within 3 years, plus the 9-month allowance of SR 3.0.2, as measured from November 7, 2004, the date of the most recent successful tracer gas test, as stated in the February 7, 2005 letter response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

(c) The first performance of the periodic measurement of CRE pressure, Specification 5.5.18.d, shall be within 18 months, plus the 138 days allowed by SR 3.0.2, as measured from the date of the most recent successful pressure measurement test, or within 138 days if not performed previously.

Amended License Condition:

This license condition is deleted in its entirety.

The licensee states in its February 23, 2017, letter that the first performances of the CRE habitability surveillance, assessment, and measurement as required by the license condition items (a), (b), and (c) respectively are complete. The NRC staff finds that after the completion of the first performances of (a), (b), and (c) the license condition is no longer necessary.

Therefore, the NRC staff determines that 10 CFR 50.50 is met with this license condition deleted.

Byron, Unit No. 1, Appendix C, Additional Conditions, Facility Operating License No. NPF-37 Current License Condition:

Amendment Number 151 Additional Condition Upon implementation of Amendment No. 151 adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 3.7.10.4, in accordance with TS 5.5.18.c.(i), the assessment of CRE habitability as required by Specification 5.5.18.c.(ii), and the measurement of CRE pressure as required by Specification 5.5.18.d, shall be considered met. Following implementation:

(a) The first performance of SR 3.7.10.4, in accordance with Specification 5.5.18.c.(i), shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from November 1, 2004, the date of the most recent successful tracer gas test, as stated in the January 31, 2005 letter response to Generic Letter 2003-01, or within the next 18 months if the time period since the most recent successful tracer gas test Is greater than 6 years.

(b) The first performance of the periodic assessment of CRE habitability, Specification 5.5.18.c.(ii), shall be within 3 years, plus the 9-month allowance of SR 3.0.2, as measured from November 1, 2004, the date of the most recent successful tracer gas test, as stated In the January 31, 2005 letter response to GL 2003-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

(c) The first performance of the periodic measurement of CRE pressure, Specification 5.5.18.d, shall be within 18 months, plus the 138 days allowed by SR 3.0.2, as measured from the date of the most recent successful pressure measurement test, or within 138 days if not performed previously.

Implementation Date With implementation of the amendment Amended License Condition:

This license condition is deleted in its entirety.

The licensee states in its February 23, 2017, letter that the first performances of the CRE habitability surveillance, assessment, and measurement as required by the license condition items (a), (b), and (c) respectively are complete. The NRC staff finds that after the completion of the first performances of (a), (b), and (c) the license condition is no longer necessary.

Therefore, the NRC staff determines that 10 CFR 50.50 is met with this license condition deleted.

Byron, Unit No. 2, Appendix C, Additional Conditions, Facility Operating License No. NPF-66 Current License Condition:

Amendment Number 151 Additional Condition Upon implementation of Amendment No. 151 adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 3.7.10.4, in accordance with TS 5.5.18.c.(i), the assessment of CRE habitability as required by Specification 5.5.18.c.(ii), and the measurement of CRE pressure as required by Specification 5.5.18.d, shall be considered met. Following implementation:

(a) The first performance of SR 3. 7.10.4, in accordance with Specification 5.5.18.c.(i), shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from November 1, 2004, the date of the most recent successful tracer gas test, as stated in the January 31, 2005 letter response to Generic Letter 2003-01, or within the next 18 months if the time period since the most recent successful tracer gas test Is greater than 6 years.

(b) The first performance of the periodic assessment of CRE habitability, Specification 5.5.18.c.(ii), shall be within 3 years, plus the 9-month allowance of SR 3.0.2, as measured from November 1, 2004, the date of the most recent successful tracer gas test, as stated In the January 31, 2005 letter response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

(c) The first performance of the periodic measurement of CRE pressure, Specification 5.5.18.d, shall be Implementation Date With implementation of the amendment within 18 months, plus the 138 days allowed by SR 3.0.2, as measured from the date of the most recent successful pressure measurement test, or within 138 days if not performed previously.

Amended License Condition:

This license condition is deleted in its entirety.

The licensee states in its February 23, 2017, letter that the first performances of the CRE habitability surveillance, assessment, and measurement as required by the license condition items (a), (b), and (c) respectively, are complete. The NRC staff finds that after the completion of the first performances of (a), (b), and (c) the license condition is no longer necessary.

Therefore, the NRC staff determines that 10 CFR 50.50 is met with this license condition deleted.

3.2 Technical Specifications 3.2.1 Braidwood, Units 1 and 2 Pages 2.0-1, 3.0-4, 3.0-6. and 4.0-2 Replace the current bottom single line with a double line, one space below the last text.

The NRC staff finds that this is an administrative change with no impact on TS implementation.

Based on the above, the NRC staff determined that this change is acceptable and the requirements of 10 CFR 50.36 continue to be met.

Page 2.0-1 Existing TS 2.1.1.3 2.1.1.3 In MODES 1 and 2, the peak fuel centerline temperature shall be maintained as follows:

a. < 5080 °F decreasing by 58 °F per 10,000 MWD/MTU burnup for Westinghouse fuel,

b. < 5173 °F decreasing by 65 °F per 10,000 MWD/MTU burnup for AREVA NP fuel (Unit 1 only), and

c.

< 5189 °F decreasing by 65 °F per 10,000 MWD/MTU burnup for AREVA NP fuel containing Gadolinia (Unit 1 only).

Amended TS 2.1.1.3 2.1.1.3 In MODES 1 and 2, the peak fuel centerline temperature shall be maintained

< 5080 °F decreasing by 58 °F per 10, 000 MWD/MTU burn up for Westinghouse fuel.

In its letter dated February 23, 2017, the licensee stated that AREVA NP fuel is no longer in the reactor and won't be reinserted. Based on the licensee's statement, the NRC staff finds that the TS specifying the peak fuel centerline temperature limits for AREVA NP fuel (TS 2.1.1.3.b and TS 2.1.1.2.c) are no longer necessary. Based on the above, the NRC staff determined that the change is acceptable and the requirements of 10 CFR 50.36 continue to be met.

Page 3.1.7-2 Existing SR 3.1.7.1 FREQUENCY:

Prior to criticality after each removal of the reactor head.

Amended SR 3.1.7.1 FREQUENCY:

Prior to criticality after each removal of the reactor head Removal of the period from the FREQUENCY statement does not change the meaning or implementation of the SR and makes the statement consistent other SRs. The NRC staff finds that this is an administrative change with no impact on TS implementation. Based on the above, the NRC staff determined that this change is acceptable and the requirements of 10 CFR 50.36 continue to be met.

Page 3.3.1-13 In the header SURVEILLANCE REQUIREMENTS, delete the text "(continued)". In its letter dated February 23, 2017, the licensee states that because SR 3.3.1.8 has the text "(continued)"

it is not appropriate to have it in the header. The NRC staff finds that this is an administrative change with no impact on TS implementation. Based on the above, the NRC staff determined that this change is acceptable and the requirements of 10 CFR 50.36 continue to be met.

Page 3.4.15-3 and 3.8.4-2 At the top of the pages above the top line add the header:

CONDITION REQUIRED ACTION COMPLETION TIME The NRC staff finds that this is an administrative change that makes this page consistent with other TS pages and has no impact on TS implementation. Based on the above, the NRC staff determined that this change is acceptable and the requirements of 10 CFR 50.36 continue to be met.

Page 3.7.1-1 The page is amended to add the text "(continued)" below the table at the bottom right to indicate continuation of the table on the next page. The NRC staff finds that this is an administrative change that makes this page consistent with other TS pages that are continued and has no impact on TS implementation. Based on the above, the NRC staff determined that this change is acceptable and the requirements of 10 CFR 50.36 continue to be met.

Unit 2 Page 3.7.8-2 Existing TS 3. 7.8.A ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One unit-specific SX

NOTES---------

[essential service

1. Enter applicable Conditions water] train and Required Actions of LCO inoperable.

[licensing condition for operation] 3.8.1, "AC

[alternating current] Sources Operating," for Emergency Diesel Generator made inoperable by SX.

2. Enter applicable Conditions and Required Actions of LCO 3.4.6, "RCS [reactor coolant system] Loops-MODE 4," for Residual Heat Removal loops made inoperable by SX.

A.1

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

Not applicable to Unit 2 during repair of the 2A SX pump during the one-time Unit 2 planned SX System outage to be completed 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months no later than January 23, 2017.

Restore unit-specific SX train to OPERABLE status.

OR A. 2

NOTE--------

Applicable to Unit 2 during repair of the 2A SX pump during the one-time planned SX System outage to be completed no later than January 23, 2017.

Allowance of the extended completion time is contingent on meeting the compensatory measures described in EGC 200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months submittal letter RS-16-197.

Restore unit-specific SX train to OPERABLE status.

Amended TS 3.7.8.A ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A

One unit-specific SX A.1

NOTES---------

train inoperable.

1. Enter applicable Conditions and Required Actions of LCO 3.8.1 I "AC Sources Operating,"

for Emergency Diesel Generator made inoperable by SX.

2. Enter applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops-MODE 4," for Residual Heat Removal loops made inoperable by SX.

Restore unit-specific SX 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months train to OPERABLE status.

The removed note after A.1 is only applicable during the one-time planned SX System outage to be completed no later than January 23, 2017. The entire TS for A.2, including the completion time of 200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months is only applicable during the one-time planned SX System outage to be completed no later than January 23, 2017. Since the January 23, 2017, date is now expired, the note after A.1 and the entire A.2 TS can be removed without any impact to implementation of the TS.

Since the Unit 2 TSs are now the same as Unit 1 TS, Unit 1 and Unit 2 TS will be specified on a common page which is consistent with the rest of the TS.

Based on the above, the NRC staff determined that this change is acceptable and the requirements of 10 CFR 50.36 continue to be met.

Page 3.7.10-3 Existing TS Condition E CONDITION E. Two VC [ventilation concentration] Filtration System Trains inoperable in MODE 5 or 6, or during movement of irradiated fuel assemblies.

One or more VC Filtration System trains inoperable due to an inoperable CRE boundary in MODE 5 or 6, or during movement of irradiated fuel assemblies. Amended TS Condition E CONDITION E. Two VC Filtration System Trains inoperable in MODE 5 or 6, or during movement of irradiated fuel assemblies.

One or more VC Filtration System trains inoperable due to an inoperable CRE boundary in MODE 5 or 6, or during movement of irradiated fuel assemblies.

The formatting change for the "OR" connector does not impact implementation of the TS and makes it consistent with the TS Section 1.2, Logical Connectors, BACKGROUND, which states,

"... the logical connector is left justified with the statement of the Condition... "

Based on the above, the NRC staff determined that this change is acceptable and the requirements of 10 CFR 50.36 continue to be met.

Pages 3.8.5-2, and 3.8.8-2 In the header ACTIONS, delete the text "(continued)." In its letter dated February 23, 2017, the licensee states that because the CONDITION column has the text "(continued)" it is not appropriate to have it in the header. The NRC staff finds that this is an administrative change with no impact on TS implementation. Based on the above, the NRC staff determined that this change is acceptable and the requirements of 10 CFR 50.36 continue to be met.

Page 4.0-1 Existing TS 4.2.1, Fuel Assemblies The reactor shall contain 193 fuel assemblies. Each assembly, with exceptions as noted below, shall consist of a matrix of Zircaloy, ZIRLO, or Optimized ZIRLO' clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods or vacancies for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

Up to 8 AREVA NP Advanced Mark-BW(A) fuel assemblies containing MS alloy may be placed in nonlimiting Unit 1 core regions for evaluation during Cycles 1 S, 16, and 17.

Amended TS 4.2.1, Fuel Assemblies The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy, ZIRLO, or Optimized ZIRLO' clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods or vacancies for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

The removed text "with exceptions as noted below," and the removed exception that specifies that, "Up to 8 AREVA NP Advanced Mark-BW(A) fuel assemblies containing MS alloy may be placed in nonlimiting Unit 1 core regions for evaluation during Cycles 1 S, 16, and 17," is only applicable during cycles 1 S, 16, and 17. In its February 23, 2017, letter the licensee stated that Cycles 1S, 16, and 17 are complete and the AREVA NP Advanced Mark-BW(A) fuel assemblies containing MS alloy are no longer in the core.

Based on the above, the NRC staff determined that this change is acceptable and the requirements of 10 CFR S0.36 continue to be met.

Page 4.0-2 Existing heading:

Amended heading:

DESIGN FEATURES (continued) 4.0 DESIGN FEATURES Adding 4.0 in front of the heading makes the page consistent with the rest of the TS pages. The text (continued) is removed because the TS Section concluded on the previous page. The NRC staff determined that these changes are administrative in nature and don't impact TS implementation.

Based on the above, the NRC staff determined that this change is acceptable and the requirements of 10 CFR S0.36 continue to be met.

Page S.S-21 Existing first paragraph of TS S.S.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 SO.S4(o) and 10 CFR SO, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, September 199S and NEI [Nuclear Energy Institute] 94-01, Revision 0, as modified by the following exceptions:

1. NEI 94 1995, Section 9.2.3: The first Unit 1 Type A test performed after the October 5, 1998 Type A test shall be performed no later than October 5, 2013.

2. NEI 94 1995, Section 9.2.3: In support of the Spring 2014 refueling outage, Unit 2 shall be placed in a MODE of operation where containment is not required to be OPERABLE in accordance with Technical Specification 3.6.1, "Containment," no later than May 4, 2014. The first Unit 2 Type A test performed after the May 4, 1999 Type A test shall be performed prior to entering MODE 4 at the start of Unit 2, Cycle 18.

Amended first paragraph of TS 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(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, September 1995 and NEI 94-01, Revision 0.

In its letter dated February 23, 2017, the licensee states that the Type A tests were performed on September 27, 2013, and May 13, 2014, for Braidwood, Units 1 and 2, respectively. These were the first Type A tests performed after the October 5, 1998, Type A test and the May 4, 1999, Type A test for Braidwood, Units 1 and 2, respectively. The NRC staff finds that with the completion of these tests, the removed text, "as modified by the following exceptions:" and exceptions 1 and 2 are no longer applicable.

Based on the above, the NRC staff determined that this change is acceptable and the requirements of 1 O CFR 50.36 continue to be met.

3.2.2 Byron, Unit Nos. 1 and 2 Pages 2.0-1, 3.0-4, 3.0-6, and 4.0-2 Replace the current bottom single line with a double line, one space below the last text.

The NRC staff finds that this is an administrative change with no impact on TS implementation.

Based on the above, the NRC staff determined that this change is acceptable and the requirements of 10 CFR 50.36 continue to be met.

Page 3.1.7-2 Existing SR 3.1.7.1 FREQUENCY:

Prior to criticality after each removal of the reactor head.

Amended SR 3.1.7.1 FREQUENCY:

Prior to criticality after each removal of the reactor head Removal of the period from the FREQUENCY statement does not change the meaning or implementation of the SR and makes the statement consistent with other SRs. The NRC staff finds that this is an administrative change with no impact on TS implementation. Based on the above, the NRC staff determined that this change is acceptable and the requirements of 10 CFR 50.36 continue to be met.

Page 3.3.1-14 In the header SURVEILLANCE REQUIREMENTS, delete the text "(continued)". In its letter dated February 23, 2017, the licensee states that because SR 3.3.1.8 has the text "(continued)"

it is not appropriate to have it in the header. The NRC staff finds that this is an administrative change with no impact on TS implementation. Based on the above, the NRC staff determined that this change is acceptable and the requirements of 1 O CFR 50.36 continue to be met.

Page 3.4.15-3. 3. 7.1-2. and 3.8.4-2 At the top of the pages above the top line add the header:

CONDITION REQUIRED ACTION COMPLETION TIME The NRC staff finds that this is an administrative change that makes this page consistent with other TS pages and has no impact on TS implementation. Based on the above, the NRC staff determined that this change is acceptable and the requirements of 10 CFR 50.36 continue to be met.

Page 3.7.1-1 The page is amended to add the text "(continued)" below the table at the bottom right to indicate continuation of the table on the next page. The NRC staff finds that this is an administrative change that makes this page consistent with other TS pages that are continued and has no impact on TS implementation. Based on the above, the NRC staff determined that this change is acceptable and the requirements of 10 CFR 50.36 continue to be met.

Pages 3.7.8-1 and 3.7.8-2 Existing TS 3.7.8 ACTIONS CONDITIONS A.

N 0 TE-------

Not applicable to Unit 1 during replacement of the SX suction isolation valves (i.e., ISX001A and 2SX001A) during Unit 2 Refueling 15 while Unit 2 is in MODE 5, 6, or defueled.

One unit-specific SX train inoperable A.1 REQUIRED ACTION COMPLETION TIME

- - - - -NOTES- - - - - -

1. Enter applicable Conditions and Required Actions of LCO 3.8.1, "AC Sources-Operating," for Emergency Diesel Generator made inoperable by SX.

2. Enter applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops-MODE 4,"

for Residual Heat Removal loops made inoperable by SX.

Restore unit-specific SX 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months train to OPERABLE status.

B.

NOTE-------

B.1 --------NOTES-------

Only applicable to Unit 1

1. Enter applicable during replacement of Conditions and Required the SX suction isolation Actions of LCO 3.8.1, valves (i.e., 1SX001A "AC and 2SX001A) during Sources-Operating," for Unit 2 Refueling 15 Emergency Diesel while Unit 2 is in Generator made MODE 5, 6, or defueled.

inoperable by SX.

2. Enter applicable One unit-specific SX Conditions and Required train inoperable.

Actions of LCO 3.4.6, "RCS Loops-MODE 4,"

for Residual Heat Removal loops made inoperable by SX.

Restore unit-specific SX train to OPERABLE status.

144 hours0.00167 days 0.04 hours 2.380952e-4 weeks 5.4792e-5 months C. Opposite-unit SX train C.1 Restore opposite-unit SX 7 days inoperable.

train to OPERABLE status.

D. Required Action and D.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A, B or AND C not met.

D.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Amended TS 3.7.8 ACTIONS CONDITIONS REQUIRED ACTION COMPLETION TIME A One unit-specific SX train A.1

- - - - -NOTES- - - - - -

inoperable

1. Enter applicable Conditions and Required Actions of LCO 3.8.1, "AC Sources-Operating,"

for Emergency Diesel Generator made inoperable by SX.

2. Enter applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops-MODE 4," for Residual Heat Removal loops made inoperable by SX.

Restore unit-specific SX 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months train to OPERABLE status.

B. Opposite-unit SX train B.1 Restore opposite-unit SX 7 days inoperable.

train to OPERABLE status.

C. Required Action and C.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A or B AND not met.

C.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months The removed note after CONDITION A is only applicable during replacement of the SX suction isolation valves (i.e., 1 SX001A and 2SX001A) during Unit No. 2 refueling outage 15. The entire TS for CONDITION B., including the completion time of 144 hours0.00167 days 0.04 hours 2.380952e-4 weeks 5.4792e-5 months is only applicable during replacement of the SX suction isolation valves (i.e., 1 SX001A and 2SX001A) during Unit No. 2 refueling outage 15. In its letter dated February 23, 2017, the licensee stated that the Byron, Unit No. 2, refueling outage 15 is complete. Since the Byron, Unit No. 2, refueling outage 15 is complete, the NRC staff finds that the note after CONDITION A and the entire CONDITION B TS can be removed without any impact to implementation of the TS. The NRC staff finds that renumbering of the CONDITIONS and REQUIRED ACTIONs because the entire CONDITION B TS is removed is an administrative change that has no impact on the TS.

Based on the above, the NRC staff determined that this change is acceptable and the requirements of 10 CFR 50.36 continue to be met.

Page 3.7.9-2 Existing TS Condition C CONDITION C. Outside air wet bulb temperature> 76°F.

Any electrical division not capable of providing power to at least one OPERABLE SXCTfan.

Amended TS Condition C CONDITION C. Outside air wet bulb temperature> 76°F.

Any electrical division not capable of providing power to at least one OPERABLE SXCT fan.

The formatting change for the "AND" connector does not impact implementation of the TS and makes it consistent with the TS, Section 1.2, Logical Connectors, BACKGROUND, which states,

"... the logical connector is left justified with the statement of the Condition... "

Based on the above, the NRC staff determined that this change is acceptable and the requirements of 10 CFR 50.36 continue to be met.

Page 3. 7.10-3 Existing TS Condition E CONDITION E. Two VC Filtration System Trains inoperable in MODE 5 or 6, or during movement of irradiated fuel assemblies.

One or more VC Filtration System trains inoperable due to an inoperable CRE boundary in MODE 5 or 6, or during movement of irradiated fuel assemblies.

Amended TS Condition E CONDITION E. Two VC Filtration System Trains inoperable in MODE 5 or 6, or during movement of irradiated fuel assemblies.

One or more VC Filtration System trains inoperable due to an inoperable CRE boundary in MODE 5 or 6, or during movement of irradiated fuel assemblies.

The formatting change for the "OR" connector does not impact implementation of the TS and makes it consistent with the TS Section 1.2, Logical Connectors, BACKGROUND, which states,

"... the logical connector is left justified with the statement of the Condition... "

Based on the above, the NRC staff determined that this change is acceptable and the requirements of 1 O CFR 50.36 continue to be met.

Page 3.7.10-4 Existing SR 3.7.10.4 FREQUENCY requirement:

In accordance with the Control Room Envelope Habitability Program.

Amended SR 3. 7.10.4 FREQUENCY requirement:

In accordance with the Control Room Envelope Habitability Program The removed period"." At the end of the requirement does not impact implementation of the SR.

Based on the above, the NRC staff determined that this change is acceptable and the requirements of 10 CFR 50.36 continue to be met.

Page 3.8.5-2 In the header ACTIONS, delete the text "(continued)". In its letter dated February 23, 2017, the licensee states that because the CONDITION column has the text "(continued)" it is not appropriate to have it in the header. The NRC staff finds that this is an administrative change with no impact on TS implementation. Based on the above, the NRC staff determined that this change is acceptable and the requirements of 10 CFR 50.36 continue to be met.

Page 4.0-2 Existing heading:

DESIGN FEATURES (continued)

Amended heading:

4.0 DESIGN FEATURES Adding 4.0 in front of the heading makes the page consistent with the rest of the TS pages. The text (continued) is removed because the TS Section concluded on the previous page. The NRC staff determined that these changes are administrative in nature and don't impact TS implementation.

Based on the above, the NRC staff determined that this change is acceptable and the requirements of 10 CFR 50.36 continue to be met.

Move the following from page 4.0-3 to page 4.0-2 after 4.3.1:

2.3.2 Drainage The spent fuel pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 410 ft. 0 inches.

2.3.3 Capacity The spent fuel pool is designed and shall be maintained with a storage capacity limited to no more than 2984 fuel assemblies.

The NRC staff finds that Moving 4.3.2 and 4.3.3 to page 4.0-2 does not change the sequence of the TS and does not impact implementation of the TS.

Based on the above, the NRC staff determined that this change is acceptable and the requirements of 1 O CFR 50.36 continue to be met.

Page 4.0.3 Delete the entire page since 4.3.2 and 4.3.3 were moved to the previous page and, therefore, the page has no requirements.

The NRC staff finds that this is an administrative repagination and doesn't impact TS implementation.

Based on the above, the NRC staff determined that this change is acceptable and the requirements of 10 CFR 50.36 continue to be met.

Page 5.5-21 Existing first paragraph of TS 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(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, September 1995 and NEI 94-01, Revision 0, as modified by the following exceptions:

1. NEI 94 1995, Section 9.2.3: The first Unit 1 Type A test performed after the February 19, 1998 Type A test shall be performed no later than February 19, 2013.

2. NEI 94 1995, Section 9.2.3: The first Unit 2 Type A test performed after the November 2, 1999 Type A test shall be performed no later than November 2, 2014.

Amended first paragraph of TS 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(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, September 1995 and NEI 94-01, Revision 0.

In its letter dated February 23, 2017, the licensee states that the Type A tests were performed on September 27, 2012, and September 26, 2014, for Byron, Unit Nos. 1 and 2, respectively.

These were the first Type A tests performed after the February 19, 1998, Type A test and the November 2, 1999, Type A test for Byron, Unit Nos. 1 and 2, respectively. The NRC staff finds that with the completion of these tests, the removed text, "as modified by the following exceptions:" and the exceptions for Unit Nos. 1 and 2 are no longer applicable.

Based on the above, the NRC staff determined that this change is acceptable and the requirements of 10 CFR 50.36 continue to be met.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Illinois State official was notified of the proposed issuance of the amendment on April 26, 2017. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment relates to changes in the format of the license or otherwise makes editorial, corrective, or minor revisions. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(10). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

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

Principal Contributor: Joel S. Wiebe Date of issuance: July 5, 2017

ML17088A703 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NAME JWiebe SRohrer DATE 4/20/2017 4/20/2017 OFFICE NRR/DSS/STSB/BC OGC NAME JWhitman(A)

RNorwood/NLO DATE 4/21/2017 5/17/2017 NRR/DSS/SBPB/BC RDennig 4/20/2017 NRR/DORL/LPL3/BC DWrona 7/5/2017

v t e Letter Sequence Approval
CAC:MF9338, Control Room Habitability (Approved, Closed) CAC:MF9339, Control Room Habitability (Approved, Closed) CAC:MF9340, Control Room Habitability (Approved, Closed) CAC:MF9341, Control Room Habitability (Approved, Closed)
Initiation Request Acceptance... Administration Questions Answers 29 June 2017 Results Approval, Approval Other:
MONTHYEAR2017-06-29 [Table View]

ML17088A703

Text

SUBJECT:

Dear Mr. Hanson:

Enclosures:

Attachment:

Attachment:

Attachment:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

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