ML22220A260

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Issuance of Amendment Nos. 322, 345, and 305 Regarding Adoption of TSTF Traveler TSTF-205-A, Rev. 3, and TSTF-563-A
ML22220A260
Person / Time
Site: Browns Ferry  
Issue date: 11/21/2022
From: Kimberly Green
Plant Licensing Branch II
To: Jim Barstow
Tennessee Valley Authority
Green K
References
EPID L-2021-LLA-0217
Download: ML22220A260 (93)
v • d • e


Text

November 21, 2022 Mr. James Barstow Vice President, Nuclear Regulatory Affairs and Support Services Tennessee Valley Authority 1101 Market Street, LP 4A-C Chattanooga, TN 37402-2801

SUBJECT:

BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2, AND 3 ISSUANCE OF AMENDMENT NOS. 322, 345, AND 305 REGARDING ADOPTION OF TECHNICAL SPECIFICATION TASK FORCE (TSTF) TRAVELER TSTF-205-A, REVISION 3, REVISION OF CHANNEL CALIBRATION, CHANNEL FUNCTIONAL TEST, AND RELATED DEFINITIONS, AND TSTF-563-A, REVISE INSTRUMENT TESTING DEFINITIONS TO INCORPORATE THE SURVEILLANCE FREQUENCY CONTROL PROGRAM (EPID L-2021-LLA-0217)

Dear Mr. Barstow:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment Nos. 322, 345, and 305 to Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 for the Browns Ferry Nuclear Plant, Units 1, 2, and 3, respectively. These amendments are in response to your application dated December 3, 2021.

The amendments revise a few instrument testing and calibration definitions in the Browns Ferry Nuclear Plant, Units 1, 2, and 3, technical specifications (TSs), and incorporate the surveillance frequency control program into a couple of these definitions. The amendments are based on TS Task Force (TSTF) Traveler TSTF-205-A, Revision 3, Revision of Channel Calibration, Channel Functional Test, and Related Definitions, and TSTF-563-A, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program. The amendments also rescind the consolidation of several previously approved surveillance requirements approved in Amendment Nos. 315, 338, and 298.

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

Sincerely,

/RA/

Kimberly J. Green, Senior Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-259, 50-260, and 50-296

Enclosures:

1. Amendment No. 322 to DPR-33
2. Amendment No. 345 to DPR-52
3. Amendment No. 305 to DPR-68
4. Safety Evaluation cc: Listserv

TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 322 Renewed License No. DPR-33

1.

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

A.

The application for amendment by the Tennessee Valley Authority (the licensee) dated December 3, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR)

Chapter I; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-33 is hereby amended to read, in part, as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 322, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: November 21, 2022 David J.

Wrona Digitally signed by David J. Wrona Date: 2022.11.21 12:56:12 -05'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 322 BROWNS FERRY NUCLEAR PLANT, UNIT 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-33 DOCKET NO. 50-259 Replace page 3 of Renewed Facility Operating License No. DPR-33 with the attached revised page 3. The revised page is identified by amendment number and contains a marginal line indicating the areas of change.

Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 1.1-1 1.1-1 1.1-2 1.1-2 1.1-5 1.1-5 3.3-4 3.3-4 3.3-5 3.3-5 3.3-7 3.3-7 3.3-8 3.3-8 3.3-11 3.3-11 3.3-14 3.3-14 3.3-25 3.3-25 3.3-28 3.3-28 3.3-41 3.3-41 3.3-42 3.3-42 3.3-43 3.3-43 3.3-44 3.3-44 3.3-45 3.3-45 3.3-46 3.3-46 3.3-47 3.3-47 3.3-68 3.3-68 3.3-69 3.3-69 3.3-73 3.3-73 3.3-74 3.3-74

BFN-UNIT 1 Renewed License No. DPR-33 Amendment No. 322 (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 equipment and instrument calibration or associated with radioactive apparatus or components; (5)

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

C.

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

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

(1)

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 3952 megawatts thermal.

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 322, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

For Surveillance Requirements (SRs) that are new in Amendment 234 to Facility Operating License DPR-33, the first performance is due at the end of the first surveillance interval that begins at implementation of the Amendment 234. For SRs that existed prior to Amendment 234, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the surveillance was last performed prior to implementation of Amendment 234.

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TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 345 Renewed License No. DPR-52

1.

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

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated December 3, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR)

Chapter I; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-52 is hereby amended to read, in part, as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 345, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: November 21, 2022 David J.

Wrona Digitally signed by David J. Wrona Date: 2022.11.21 12:56:51 -05'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 345 BROWNS FERRY NUCLEAR PLANT, UNIT 2 RENEWED FACILITY OPERATING LICENSE NO. DPR-52 DOCKET NO. 50-260 Replace page 3 of Renewed Facility Operating License No. DPR-52 with the attached revised page 3. The revised page is identified by amendment number and contains a marginal line indicating the areas of change.

Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 1.1-1 1.1-1 1.1-2 1.1-2 1.1-5 1.1-5 3.3-5 3.3-5 3.3-6 3.3-6 3.3-8 3.3-8 3.3-9 3.3-9 3.3-12 3.3-12 3.3-15 3.3-15 3.3-26 3.3-26 3.3-29 3.3-29 3.3-42 3.3-42 3.3-43 3.3-43 3.3-44 3.3-44 3.3-45 3.3-45 3.3-46 3.3-46 3.3-47 3.3-47 3.3-48 3.3-48 3.3-69 3.3-69 3.3-70 3.3-70 3.3-74 3.3-74 3.3-75 3.3-75

BFN-UNIT 2 Renewed License No. DPR-52 Amendment No. 345 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 equipment and instrument calibration or associated with radioactive apparatus or components; (5)

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

C.

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

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

(1)

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 3952 megawatts thermal.

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 345, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

For Surveillance Requirements (SRs) that are new in Amendment 253 to Facility Operating License DPR-52, the first performance is due at the end of the first surveillance interval that begins at implementation of the Amendment 253. For SRs that existed prior to Amendment 253, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the surveillance was last performed prior to implementation of Amendment 253.

3)

The licensee is authorized to relocate certain requirements included in Appendix A and the former Appendix B to licensee-controlled documents.

Implementation of this amendment shall include the relocation of these requirements to the appropriate documents, as described in the licensee's

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TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 305 Renewed License No. DPR-68

1.

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

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated December 3, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR)

Chapter I; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-68 is hereby amended to read, in part, as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 305, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: November 21, 2022 David J.

Wrona Digitally signed by David J. Wrona Date: 2022.11.21 12:57:28 -05'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 305 BROWNS FERRY NUCLEAR PLANT, UNIT 3 RENEWED FACILITY OPERATING LICENSE NO. DPR-68 DOCKET NO. 50-296 Replace page 3 of Renewed Facility Operating License No. DPR-68 with the attached revised page 3. The revised page is identified by amendment number and contains a marginal line indicating the areas of change.

Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 1.1-1 1.1-1 1.1-2 1.1-2 1.1-5 1.1-5 3.3-5 3.3-5 3.3-6 3.3-6 3.3-8 3.3-8 3.3-9 3.3-9 3.3-12 3.3-12 3.3-15 3.3-15 3.3-26 3.3-26 3.3-29 3.3-29 3.3-42 3.3-42 3.3-43 3.3-43 3.3-44 3.3-44 3.3-45 3.3-45 3.3-46 3.3-46 3.3-47 3.3-47 3.3-48 3.3-48 3.3-69 3.3-69 3.3-70 3.3-70 3.3-74 3.3-74 3.3-75 3.3-75

BFN-UNIT 3 Renewed License D

No. PR-68 Amendment No. 305 (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 equipment and instrument calibration or associated with radioactive apparatus or components; (5)

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

C.

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

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

(1)

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 3952 megawatts thermal.

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 305, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

For Surveillance Requirements (SRs) that are new in Amendment 212 to Facility Operating License DPR-68, the first performance is due at the end of the first surveillance interval that begins at implementation of the Amendment 212. For SRs that existed prior to Amendment 212, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the surveillance was last performed prior to implementation of Amendment 212.

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ECCS Instrumentation 3.3.5.1 BFN-UNIT 3 3.3-43 Table 3.3.5.1-1 (page 1 of 6)

Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE 1.

Core Spray System a.

Reactor Vessel Water Level Low Low Low, Level 1(f) 1,2,3 4(b)

B SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.5 SR 3.3.5.1.6 t 398 inches above vessel zero b.

Drywell Pressure High(f) 1,2,3 4(b)

B SR 3.3.5.1.2 SR 3.3.5.1.5 SR 3.3.5.1.6 d 2.5 psig c.

Reactor Steam Dome Pressure Low (Injection Permissive and ECCS Initiation)(f) 1,2,3 4(b) 2 per trip system C

SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 t 435 psig and d 465 psig d.

Core Spray Pump Discharge Flow Low (Bypass) 1,2,3 2

1 per subsystem E

SR 3.3.5.1.2 SR 3.3.5.1.5 t 1647 gpm and d 2910 gpm e.

Core Spray Pump Start Time Delay Relay Pumps A,B,C,D (with diesel power) 1,2,3 4

1 per pump C

SR 3.3.5.1.5 SR 3.3.5.1.6 t 6 seconds and d 8 seconds Pump A (with normal power) 1,2,3 1

C SR 3.3.5.1.5 SR 3.3.5.1.6 t 0 seconds and d 1 second Pump B (with normal power) 1,2,3 1

C SR 3.3.5.1.5 SR 3.3.5.1.6 t 6 seconds and d 8 seconds (continued)

D 'HOHWHG

E Channels affect Common Accident Signal Logic. Refer to LCO 3.8.1, "AC Sources - Operating."

(f)

During instrument calibrations, if the As Found channel setpoint is conservative with respect to the Allowable Value but outside its acceptable As Found band as defined by its associated Surveillance Requirement procedure, then there shall be an initial determination to ensure confidence that the channel can perform as required before returning the channel to service in accordance with the Surveillance. If the As Found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable.

Prior to returning a channel to service, the instrument channel setpoint shall be calibrated to a value that is within the acceptable As Left tolerance of the setpoint; otherwise, the channel shall be declared inoperable.

The nominal Trip Setpoint shall be specified on design output documentation which is incorporated by reference in the Updated Final Safety Analysis Report. The methodology used to determine the nominal Trip Setpoint, the predefined As Found Tolerance, and the As Left Tolerance band, and a listing of the setpoint design output documentation shall be specified in Chapter 7 of the Updated Final Safety Analysis Report.

Amendment No. 213, 254, 



ECCS Instrumentation 3.3.5.1 BFN-UNIT 3 3.3-44 Table 3.3.5.1-1 (page 2 of 6)

Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE 1.

Core Spray System (continued) e.

Core Spray Pump Start Time Delay Relay (continued)

Pump C (with normal power) 1,2,3 1

C SR 3.3.5.1.5 SR 3.3.5.1.6 t 12 seconds and d 16 seconds Pump D (with normal power) 1,2,3 1

C SR 3.3.5.1.5 SR 3.3.5.1.6 t 18 seconds and d 24 seconds 2.

Low Pressure Coolant Injection (LPCI) System a.

Reactor Vessel Water Level Low Low Low, Level 1(f) 1,2,3 4

B SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.5 SR 3.3.5.1.6 t 398 inches above vessel zero b.

Drywell Pressure High(f) 1,2,3 4

B SR 3.3.5.1.2 SR 3.3.5.1.5 SR 3.3.5.1.6 d 2.5 psig c.

Reactor Steam Dome Pressure Low (Injection Permissive and ECCS Initiation)(f) 1,2,3 4

C SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 t 435 psig and d 465 psig (continued)

D 'HOHWHG

E Deleted.

(f)

During instrument calibrations, if the As Found channel setpoint is conservative with respect to the Allowable Value but outside its acceptable As Found band as defined by its associated Surveillance Requirement procedure, then there shall be an initial determination to ensure confidence that the channel can perform as required before returning the channel to service in accordance with the Surveillance. If the As Found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable.

Prior to returning a channel to service, the instrument channel setpoint shall be calibrated to a value that is within the acceptable As Left tolerance of the setpoint; otherwise, the channel shall be declared inoperable.

The nominal Trip Setpoint shall be specified on design output documentation which is incorporated by reference in the Updated Final Safety Analysis Report. The methodology used to determine the nominal Trip Setpoint, the predefined As Found Tolerance, and the As Left Tolerance band, and a listing of the setpoint design output documentation shall be specified in Chapter 7 of the Updated Final Safety Analysis Report.

Amendment No. 213, 254, 



ECCS Instrumentation 3.3.5.1 BFN-UNIT 3 3.3-45 Table 3.3.5.1-1 (page 3 of 6)

Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE 2.

LPCI System (continued) d.

Reactor Steam Dome Pressure Low (Recirculation Discharge Valve Permissive)(f) 1(c),2(c),

3(c) 4 C

SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 t 215 psig and d 245 psig e.

Reactor Vessel Water Level Level 0 1,2,3 2

1 per subsystem B

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.5 SR 3.3.5.1.6 t 312 5/16 inches above vessel zero f.

Low Pressure Coolant Injection Pump Start Time Delay Relay Pump A,B,C,D (with diesel power) 1,2,3, 4

C SR 3.3.5.1.5 SR 3.3.5.1.6 t 0 seconds and d 1 second Pump A (with normal power) 1,2,3, 1

C SR 3.3.5.1.5 SR 3.3.5.1.6 t 0 seconds and d 1 second Pump B (with normal power) 1,2,3, 1

C SR 3.3.5.1.5 SR 3.3.5.1.6 t 6 seconds and d 8 seconds Pump C (with normal power) 1,2,3, 1

C SR 3.3.5.1.5 SR 3.3.5.1.6 t 12 seconds and d 16 seconds Pump D (with normal power) 1,2,3, 1

C SR 3.3.5.1.5 SR 3.3.5.1.6 t 18 seconds and d 24 seconds (continued)

(a) 'HOHWHG

(c) With associated recirculation pump discharge valve open.

(f)

During instrument calibrations, if the As Found channel setpoint is conservative with respect to the Allowable Value but outside its acceptable As Found band as defined by its associated Surveillance Requirement procedure, then there shall be an initial determination to ensure confidence that the channel can perform as required before returning the channel to service in accordance with the Surveillance. If the As Found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable.

Prior to returning a channel to service, the instrument channel setpoint shall be calibrated to a value that is within the acceptable As Left tolerance of the setpoint; otherwise, the channel shall be declared inoperable.

The nominal Trip Setpoint shall be specified on design output documentation which is incorporated by reference in the Updated Final Safety Analysis Report. The methodology used to determine the nominal Trip Setpoint, the predefined As Found Tolerance, and the As Left Tolerance band, and a listing of the setpoint design output documentation shall be specified in Chapter 7 of the Updated Final Safety Analysis Report.

Amendment No. 213, 248, 254,

298, 

ECCS Instrumentation 3.3.5.1 BFN-UNIT 3 3.3-46 Amendment No. 213, 254, 298, 

Table 3.3.5.1-1 (page 4 of 6)

Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE 3.

High Pressure Coolant Injection (HPCI) System a.

Reactor Vessel Water Level Low Low, Level 2(f) 1, 2(d), 3(d) 4 B

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.5 SR 3.3.5.1.6 t 470 inches above vessel zero b.

Drywell Pressure High(f) 1, 2(d),3(d) 4 B

SR 3.3.5.1.2 SR 3.3.5.1.5 SR 3.3.5.1.6 d 2.5 psig c.

Reactor Vessel Water Level High, Level 8 1,

2(d), 3(d) 2 C

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.5 SR 3.3.5.1.6 d 583 inches above vessel zero d.

Condensate Header Level Low 1,

2(d), 3(d) 1 D

SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.6 t Elev. 551 feet e.

Suppression Pool Water Level High 1,

2(d), 3(d) 1 D

SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.6 d 7 inches above instrument zero f.

High Pressure Coolant Injection Pump Discharge FlowLow (Bypass) 1, 2(d), 3(d) 1 E

SR 3.3.5.1.2 SR 3.3.5.1.5 SR 3.3.5.1.6 t 671 gpm 4.

Automatic Depressurization System (ADS) Trip System A a.

Reactor Vessel Water Level Low Low Low, Level 1(f) 1, 2(d), 3(d) 2 F

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.5 SR 3.3.5.1.6 t 398 inches above vessel zero (continued)

(d) With reactor steam dome pressure > 150 psig.

(f)

During instrument calibrations, if the As Found channel setpoint is conservative with respect to the Allowable Value but outside its acceptable As Found band as defined by its associated Surveillance Requirement procedure, then there shall be an initial determination to ensure confidence that the channel can perform as required before returning the channel to service in accordance with the Surveillance. If the As Found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable.

Prior to returning a channel to service, the instrument channel setpoint shall be calibrated to a value that is within the acceptable As Left tolerance of the setpoint; otherwise, the channel shall be declared inoperable.

The nominal Trip Setpoint shall be specified on design output documentation which is incorporated by reference in the Updated Final Safety Analysis Report. The methodology used to determine the nominal Trip Setpoint, the predefined As Found Tolerance, and the As Left Tolerance band, and a listing of the setpoint design output documentation shall be specified in Chapter 7 of the Updated Final Safety Analysis Report.

ECCS Instrumentation 3.3.5.1 BFN-UNIT 3 3.3-47 Amendment No. 212, 213, 219, 254, 298, 

Table 3.3.5.1-1 (page 5 of 6)

Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE 4.

ADS Trip System A (continued) b.

Drywell Pressure High(f) 1, 2(d), 3(d) 2 F

SR 3.3.5.1.2 SR 3.3.5.1.5 SR 3.3.5.1.6 d 2.5 psig c.

Automatic Depressurization System Initiation Timer 1,

2(d), 3(d) 1 G

SR 3.3.5.1.5 SR 3.3.5.1.6 d 115 seconds d.

Reactor Vessel Water Level Low, Level 3 (Confirmatory)(f) 1, 2(d), 3(d) 1 F

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.5 SR 3.3.5.1.6 t 528 inches above vessel zero e.

Core Spray Pump Discharge Pressure High 1,

2(d), 3(d) 4 G

SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.6 t 175 psig and d 195 psig f.

Low Pressure Coolant Injection Pump Discharge Pressure High 1,

2(d), 3(d) 8 G

SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.6 t 90 psig and d 110 psig g.

Automatic Depressurization System High Drywell Pressure Bypass Timer 1,

2(d), 3(d) 2 G

SR 3.3.5.1.5 SR 3.3.5.1.6 d 322 seconds 5.

ADS Trip System B a.

Reactor Vessel Water Level Low Low Low, Level 1(f) 1, 2(d), 3(d) 2 F

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.5 SR 3.3.5.1.6 t 398 inches above vessel zero (continued)

(d) With reactor steam dome pressure > 150 psig.

(f)

During instrument calibrations, if the As Found channel setpoint is conservative with respect to the Allowable Value but outside its acceptable As Found band as defined by its associated Surveillance Requirement procedure, then there shall be an initial determination to ensure confidence that the channel can perform as required before returning the channel to service in accordance with the Surveillance. If the As Found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable.

Prior to returning a channel to service, the instrument channel setpoint shall be calibrated to a value that is within the acceptable As Left tolerance of the setpoint; otherwise, the channel shall be declared inoperable.

The nominal Trip Setpoint shall be specified on design output documentation which is incorporated by reference in the Updated Final Safety Analysis Report. The methodology used to determine the nominal Trip Setpoint, the predefined As Found Tolerance, and the As Left Tolerance band, and a listing of the setpoint design output documentation shall be specified in Chapter 7 of the Updated Final Safety Analysis Report.

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 322, 345, AND 305 TO RENEWED FACILITY OPERATING LICENSE NOS. DPR-33, DPR-52, AND DPR-68 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2, AND 3 DOCKET NOS. 50-259, 50-260, AND 50-296

1.0 INTRODUCTION

By application dated December 3, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21337A227), the Tennessee Valley Authority (TVA, the licensee) submitted a license amendment request (LAR) to the U.S. Nuclear Regulatory Commission (NRC or the Commission) to revise the Browns Ferry Nuclear Plant (Browns Ferry), Units 1, 2, and 3, Technical Specifications (TSs).

The proposed amendments would revise the Browns Ferry, Units 1, 2, and 3, TS definitions for Channel Calibration, Channel Functional Test, and Logic System Functional Test. The proposed changes are based on Technical Specification Task Force (TSTF) Traveler TSTF-205-A, Revision 3, Revision of Channel Calibration, Channel Functional Test, and Related Definitions dated December 23, 1998 (ML040570179). The NRC staff approved TSTF-205, Revision 3, on January 13, 1999 (ML20199E634).

The proposed amendments would also revise the current instrumentation testing definitions of Channel Calibration and Channel Functional Test to permit determination of the appropriate frequency to perform the surveillance requirement (SR) based on the devices being tested in each step. These proposed changes are based on TSTF Traveler TSTF-563, Revision 0, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program, dated May 10, 2017 (ML17130A819). The NRC staff issued a final safety evaluation (SE) approving TSTF-563, Revision 0, on December 4, 2019 (ML18333A144).

A Surveillance Frequency Control Program (SFCP) was incorporated into the Browns Ferry, Units 1, 2, and 3, TSs in Amendment Nos. 315, 338, and 298, respectively, dated April 8, 2021 (ML21041A489). The requested changes would rescind the consolidation of several previously approved SRs approved in these amendments and restore the SRs to their prior status with the frequencies as, In accordance with the Surveillance Frequency Control Program.

2.0 REGULATORY EVALUATION

2.1 Description of SFCP and Instrument Testing The Browns Ferry, Units 1, 2, and 3 TSs require the surveillances for instrumentation channels be performed within the specified frequency, using any series of sequential, overlapping, or total channel steps. Prior amendments revised the TSs to relocate all periodic surveillance frequencies to licensee control. Changes to the relocated surveillance frequencies are made in accordance with the TS program referred to as the SFCP. The SFCP allows a new surveillance frequency to be determined for instrumentation channels, but that frequency must consider all components in the channel and applies to the entire channel.

A typical instrument channel consists of many different components, such as sensors, rack modules, and indicators. These components have different short-term and long-term performance (drift) characteristics, resulting in the potential for different calibration frequency requirements. Under the current TSs, the most limiting component calibration frequency for the channel must be chosen when a revised frequency is considered under the SFCP. As a result, all components that makeup a channel must be calibrated at a frequency equal to the channel component with the shortest (i.e., most frequent) surveillance frequency.

Some channel components, such as pressure transmitters, are very stable with respect to drift and could support a substantially longer calibration frequency than the other components in the channel. Currently, the SRs in many plants are performed in steps (e.g., a pressure sensor or transmitter is calibrated during a refueling outage and the rack signal conditioning modules are calibrated while operating at power). The proposed changes extend this concept to permit the surveillance frequency of each step to be determined under the SFCP based on the component(s) surveilled in the step instead of all components in the channel. This will allow each component to be tested at a frequency that is based on the components long-term performance characteristics rather than testing all components in a channel based on the component with the shortest surveillance frequency.

Allowing a surveillance frequency based on the components long-term performance characteristics for performing a channel calibration on each component or group of components could reduce radiation dose associated with in-place calibration of sensors, reduce wear on equipment, reduce unnecessary burden on plant staff, and reduce opportunities for calibration errors.

2.2 Proposed Changes to the Technical Specifications (TSTF-205 and TSTF-563)

Currently, the channel calibration and channel functional test may be performed by any series of sequential, overlapping, or total channel steps. The proposed changes to the TSs would revise the definitions of channel calibration and channel functional test to indicate that the step must be performed within the most limiting frequency for the components included in that step by adding the phrase, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step at the end of the last sentence of each definition.

The following paragraph denotes the proposed changes to the channel calibration definition related to TSTF-205-A. Deletions are shown in strikethrough text and additions are shown in bold italics:

A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor, alarm, display, and trip functions, and shall include all devices in the channel required for channel OPERABILITY and the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is calibrated.

The following paragraph denotes the proposed changes to the channel calibration definition related to TSTF-563. Changes are shown in bold italics:

The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

The following paragraph denotes the proposed changes to the channel functional test definition related to TSTF-205-A. Deletions are shown in strikethrough text and additions are shown in bold italics:

A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY, including required alarm, interlock, display, trip functions, and channel failure trips of all devices in the channel required for channel OPERABILITY. The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is tested.

The following paragraph denotes the proposed changes to the channel functional test definition related to TSTF-563. Changes are shown in bold italics:

The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

The following paragraph denotes the proposed changes to the Logic System Functional test definition related to TSTF-205-A. Deletions are shown in strikethrough text and additions are shown in bold italics:

A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all required logic components (i.e., all required relays and contacts, trip units, solid state logic elements, etc.) required for OPERABILITY of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY.

The various instrumentation functions in the TSs require surveillances to verify the correct functioning of the instrument channel. The proposed changes would extend the definition of instrumentation channel components to permit the surveillance frequency of each step to be determined under the SFCP based on the component(s) surveilled in the step instead of all components in the channel. This would allow each component to be tested at a frequency based on the components long-term performance characteristics.

The proposed changes in the definition for instrument testing would also allow the licensee to control the frequency of associated components being tested in each step. The SR for the overall instrumentation channel remains unchanged. The proposed changes have no effect on the design, fabrication, use, or methods of testing the instrumentation channels, and will not affect the ability of the instrumentation to perform the functions assumed in the safety analysis.

These instrumentation testing definitions state that, [t]he [test type] may be performed by means of any series of sequential, overlapping, or total channel steps. The surveillance frequency of these subsets would be established based on the characteristics of the components in the step rather than the most limiting component characteristics in the entire channel. Each of these steps would be evaluated in accordance with the SFCP.

2.3 Recission of Surveillance Requirements The licensee proposed to restore previous SR consolidations that were requested in an LAR dated March 27, 2020 (ML20087P262). The proposed changes would restore SRs 3.3.1.1.8 and 3.3.1.1.12, to perform CHANNEL FUNCTIONAL TEST; SR 3.3.1.2.3, to perform CHANNEL CHECK; SRs 3.3.5.1.4, 3.3.5.1.5, 3.3.7.1.5, and 3.3.8.1.2 to perform CHANNEL CALIBRATION; and SR 3.3.7.1.6, to perform LOGIC SYSTEM FUNCTIONAL TEST. The licensee also proposed to restore SRs 3.3.3.1.4 and 3.3.3.2.3 to perform a CHANNEL CALIBRATION for the respective instrument channels that were previously deleted. Additionally, the licensee proposed to revise the wording of SRs 3.3.3.1.3 and 3.3.3.2.2 due to the restoration of SRs 3.3.3.1.4 and 3.3.3.2.3.

2.4 Applicable Regulatory Requirements and Guidance Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(a)(1) requires, in part, each applicant for a license authorizing operation of a utilization facility to include in the application the proposed TSs.

The regulation at 10 CFR 50.36(b) requires:

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

The categories of items required to be in the TSs are provided in 10 CFR 50.36(c). One such category is SRs, which are defined in 10 CFR 50.36(c)(3) as requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

The regulation at 10 CFR 50.36(c)(5) requires TSs to include administrative controls, which are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.

Prior Browns Ferry amendments replaced specific surveillance frequencies with references to a SFCP required by TS 5.5.15 (ML21041A489). That TS requires the licensee to establish, implement, and maintain an SFCP to ensure that TS SRs are performed at intervals listed in, and controlled by, the SFCP. Technical Specification 5.5.15 also requires that changes to surveillance frequencies listed in the SFCP be made in accordance with NRC staff-approved topical report Nuclear Energy Institute (NEI) 04-10, Revision 1, Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies, Industry Guidance Document, dated April 2007 (ML071360456).

Topical report NEI 04-10, Revision 1, describes an evaluation process and a multi-disciplinary plant decision-making panel that considers the detailed evaluation of proposed surveillance frequency revisions. The evaluations are based on operating experience, test history, manufacturers recommendations, codes and standards, and other deterministic factors, in conjunction with risk insights. The evaluation considers all components being tested by the SR.

Process elements are included for determining the cumulative risk impact of the changes, updating the licensees probabilistic risk assessment (PRA) models, and for imposing corrective actions, if necessary, following implementation of a revised frequency.

The NRC staffs guidance for the review of TSs is in Chapter 16.0, Technical Specifications, of NUREG-0800, Revision 3, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition (SRP), dated March 2010 (ML100351425). As described therein, as part of the regulatory standardization effort, the NRC staff has prepared standard technical specifications (STS) for each of the LWR nuclear designs.

Accordingly, the NRC staffs review includes consideration of whether the proposed changes are consistent with the applicable reference STS (i.e., the current STS), as modified by NRC-approved Travelers. In addition, the guidance states that for facilities with TS based on previous STS, comparing the change to previous STS can help clarify the TS intent.

Regulatory Guide (RG) 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, dated January 2018 (ML17917A256), describes an acceptable risk-informed approach for assessing the nature and impact of proposed permanent licensing basis changes by considering engineering issues and applying risk insights. This RG also provides risk acceptance guidelines for evaluating the results of such evaluations.

Regulatory Guide 1.177, Revision 2, Plant-Specific, Risk-Informed Decision-Making: Technical Specifications, dated January 2021 (ML20164A034), describes an acceptable risk-informed approach specifically for assessing proposed TS changes.

Regulatory Guide 1.200, Revision 3, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, dated December 2020 (ML20238B871), describes an acceptable approach for determining the technical adequacy of PRAs.

The NRC staffs guidance for evaluating the technical basis for proposed risk-informed changes is provided in SRP Chapter 19, Section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance, dated June 2007 (ML071700658). The NRC staffs guidance for evaluating PRA technical adequacy is provided in SRP Chapter 19, Section 19.1, Revision 3, Determining the Technical Adequacy of Probabilistic Risk Assessment for Risk-Informed License Amendment Requests After Initial Fuel Load, dated September 2012 (ML12193A107). More specific review guidance related to risk-informed TS changes is provided in SRP Chapter 16, Section 16.1, Revision 1, Risk-Informed Decision Making: Technical Specifications, dated March 2007 (ML070380228),

which includes changes to surveillance test intervals (STIs) (i.e., surveillance frequencies) as part of risk-informed decision-making. Section 19.2 of the SRP references the same criteria as RG 1.177, Revision 1, and RG 1.174, Revision 2, and states that a risk-informed application should be evaluated to ensure that the proposed changes meet the following key principles:

The proposed change meets the current regulations unless it explicitly relates to a requested exemption or rule change.

The proposed change is consistent with the defense-in-depth philosophy.

The proposed change maintains sufficient safety margins.

When proposed changes result in an increase in risk associated with core damage frequency or large early release frequency, the increase(s) should be small and consistent with the intent of the Commissions Safety Goal Policy Statement.

The impact of the proposed change should be monitored using performance measurement strategies.

The NRCs, Standard Technical Specifications, General Electric Plants, BWR/4 NUREG-1433, Volume 1, Specifications, and Volume 2, Bases, Revision 5.0, dated September 2021 (ML21272A357 and ML21272A358, respectively), provide STS for General Electric BWR/4-designed reactors.

3.0 TECHNICAL EVALUATION

The NRC staff evaluated the licensees LAR to determine if the proposed changes are consistent with the regulations, guidance, and plant-specific design and licensing basis information discussed in Section 2.4 of this safety evaluation.

3.1 Adoption of TSTF-205 Traveler TSTF-205-A, Revision 3, revised the definitions for Channel Calibration, Channel Functional Test, and Logic System Functional Test in the improved STSs to remove potential ambiguity in what constitutes an acceptable test.

Section 1.1 of the Browns Ferry TSs includes definitions for instrumentation testing requirements. In accordance with 10 CFR 50.36(c)(3), Surveillance requirements, SRs are requirements related to test, calibration, or inspection needed to assure that the necessary quality of systems and components is maintained, facility operation is within limits, and that the TS limiting condition for operation is met. The definitions of Channel Calibration, Channel Functional Test, and Logic System Functional Test establish requirements for conducting testing including what the test involves, the scope of components that the test encompasses, and instructions on how the test is to be performed. The terms defined in TS Section 1.1 are referenced throughout the instrumentation TS SRs to help assure consistent performance of SRs.

In its LAR, the licensee proposed to adopt TSTF-205-A, Revision 3, TS changes to the definitions of the Channel Calibration, Channel Functional Test, and Logic System Functional Test to eliminate current ambiguity and potential misinterpretations of testing requirements.

The current definitions for instrumentation Channel Calibration, Channel Functional Test, and Logic System Functional Test use the phrases the entire channel including the sensor and alarm and/or trip functions, and including alarm and/or trip functions and channel failure trips to describe those instrument channel devices required to be included for specified tests. There is ambiguity in whether the list is inclusive of all devices that must be tested or whether the list is representative of devices to be tested. Thus, the licensee proposed to adopt changes from TSTF-205-A, Revision 3, which would replace the string of required instrument channel devices in the definitions discussed above with all devices in the channel required for channel OPERABILITY.

The revised Channel Functional Test definition does not address the method for conducting testing of all required channel devices. The NRC staff position is that a successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable channel functional test of a relay. This is acceptable because all the other required contacts of the relay are verified by other TS and non-TS tests at least once per refueling interval with applicable extensions.

This NRC staff position is incorporated into the TSTF-205-A, Revision 3, Bases for Channel Functional Tests. The licensee proposed to adopt the approved TSTF-205-A, Revision 3, Bases to clarify testing requirements by modifying the Bases of applicable surveillances to provide acceptable methods of testing.

In addition, the current TS definition for Channel Calibration specifies that testing may be performed by means of any series of sequential, overlapping or total channel steps so that the entire channel is calibrated, and the definition for Channel Functional Test specifies that testing may be performed by means of any series of sequential, overlapping or total channel steps so that the entire channel is tested. The proposed TS changes would delete the phrase such that the entire channel is calibrated from the definition of Channel Calibration and such that the entire channel is tested from the definition of Channel Functional Test to eliminate a verbatim conflict between the definition and the TSTF-205-A, Revision 3, Bases, which state a successful test to be the verification of the change of state of a single contact of the relay.

The NRC staff reviewed the changes proposed by the licensee and finds them acceptable because they are consistent with the STSs, do not result in any substantive change in operating requirements, and are consistent with 10 CFR 50.36(c)(3). These changes will provide for a consistent application of the definitions, tests, and calibrations.

3.2 Adoption of TSTF-563 Revising the frequency of a channel calibration and channel functional test instrument channel under the SFCP requires assurance that component performance characteristics, such as drift between each test, will not result in undetected instrument errors that exceed the assumptions of the safety analysis and supporting instrument loop uncertainty calculations. These requirements are consistent with the methodology described in NEI 04-10, which is required by the SFCP. The SFCP does not permit changes to the TS Allowable Values or Nominal Trip Setpoints; but allows only the surveillance frequency to be changed when determined permissible by NEI 04-10. Therefore, prior to extending the test intervals for an instrument channel component or components associated with a given calibration step, the component performance characteristics must be evaluated to verify the Allowable Value or Nominal Trip Setpoint will still be valid and to establish a firm technical basis supporting the extension. In addition, each change must be reviewed by the licensee to ensure the applicable uncertainty allowances are conservative (bounding) (e.g., sensor drift, rack drift, indicator drift).

Documentation to support the changes will be retained per the guidance in NEI 04-10.

Five key safety principles that must be evaluated before changing any surveillance frequency are identified in Section 3.0 of NEI 04-10. Principle 3 requires confirmation of the maintenance of safety margins, which, in this case, includes performance of deterministic evaluations to verify preservation of instrumentation trip setpoint and indication safety margins.

The evaluation methodology specified in NEI 04-10 also requires consideration of common-cause failure effects and monitoring of the instrument channel component performance following the frequency change to ensure channel performance is consistent with the analysis to support an extended frequency.

The method of evaluating a proposed surveillance frequency change is not dependent on the number of components in the channel. Each step needs to be evaluated to determine the acceptable surveillance frequency for that step. The proposed change to permit changing the surveillance frequency of channel component(s) does not affect the test method or evaluation method. The requirement to perform a channel calibration or channel functional test on the entire channel is not changed.

For example, an evaluation in accordance with NEI 04-10 may determine that a field sensor (e.g., a transmitter) should be calibrated every 48 months, that the rack modules should be calibrated every 30 months, and the indicators should be calibrated every 24 months. Under the current TS requirements, all devices in the channel must be calibrated every 24 months.

However, under the proposed change, sensors, rack modules, and indicators would be calibrated at the appropriate frequency for the tested devices. As required by the channel calibration definition, the test would still encompass all devices in the channel required for channel operability.

The NEI 04-10 methodology is used to evaluate surveillance frequency changes to determine if such SR extensions could be applied. Process elements are used to determine the cumulative risk impact of changes, update the PRA, and impose corrective actions, if needed, following implementation. Several steps are required by NEI 04-10, Step 7, to be evaluated prior to determining the acceptability of changes. These steps include history of surveillance tests, industry and plant specific history, impact on defense in depth, vendor recommendations, required test frequencies for the applicable codes and standards, ensuring that plant licensing basis would not be invalidated, and other factors. The NRC staff finds these measures acceptable in determining the SR extensions.

In addition, Step 16 of Section 4.0 of NEI 04-10 requires an Independent Decision-Making Panel (IDP) to review the cumulative impact of all STI changes over a period of time. This is also required by RGs 1.174 and 1.177. The IDP is composed of the site Maintenance Rule Expert Panel, Surveillance Test Coordinator, and Subject Matter Expert, who is a cognizant system manager or component engineer. Based on the above information, the NRC staff finds that the setpoint changes will be tracked in an acceptable manner.

Licensees with an SFCP may currently revise the surveillance frequency of instrumentation channels. The testing of these channels may be performed by means of any series, sequential, overlapping, or total channel steps. However, all required components in the instrumentation channel must be tested in order for the entire channel to be considered operable.

The NRC staff notes that industry practice is to perform instrument channel surveillances, such as channel calibrations and channel functional tests, using separate procedures based on the location of the components. Each of these procedures may be considered a step. The results of all these procedures are used to satisfy the SRs using the existing allowance to perform it by means of any series of sequential, overlapping, or total channel steps. The proposed changes would allow for determining an acceptable surveillance frequency for each step.

The NRC staff notes that the NEI 04-10 methodology includes the determination of whether the structure, system, and components (SSCs) affected by a proposed change to a surveillance frequency are modeled in the PRA. Where the SSC is directly or implicitly modeled, a quantitative evaluation of the risk impact may be carried out. The methodology adjusts the failure probability of the impacted SSCs based on the proposed change to the surveillance frequency. Where the SSC is not modeled in the PRA, bounding analyses are performed to characterize the impact of the proposed change to the surveillance frequency. Potential impacts on the risk analyses due to screening criteria and truncation levels are addressed by the requirements for PRA technical adequacy, consistent with the guidance contained in RG 1.200, and by sensitivity studies identified in NEI 04-10. The licensee is not proposing to change the methodology, or the acceptance criteria for extending STIs, and the licensee will need to evaluate changes in the frequency for performing each of the steps in the instrumentation surveillance test per the methodology in NEI 04-10.

Therefore, the NRC staff concludes that the proposed changes to determine an acceptable test frequency for individual steps within instrumentation channel surveillance tests is acceptable because any extended STIs will be developed within the established constraints of the SFCP and NEI 04-10.

The regulatory requirements in 10 CFR 50.36 are not specific regarding the frequency of performing surveillance tests. The proposed changes only affect the frequency of performance and do not affect the surveillance testing method or acceptance criteria. Therefore, the proposed changes are consistent with the surveillance testing requirements of 10 CFR 50.36.

PRA Acceptability The guidance in RG 1.200 states that the quality of a licensees PRA should be commensurate with the safety significance of the proposed TS change and the role the PRA plays in justifying the change. That is, the greater the change in risk or the greater the uncertainty in that risk as a result of the requested TS change, or both, the more rigor that should go into ensuring the quality of the PRA.

The NRC staff previously performed an assessment of the PRA models used to support the approved SFCP that uses NEI 04-10, using the guidance of RG 1.200 to assure that the PRA models are capable of determining the change in risk due to changes to surveillance frequencies of SSCs, using plant-specific data and models. Capability Category II of the NRC-endorsed PRA standard is the target capability level for supporting requirements for the internal events PRA for this application. Any identified deficiencies to those requirements are assessed further by the licensee to determine any impacts to proposed decreases to surveillance frequencies, including the use of sensitivity studies, where appropriate, in accordance with NEI 04-10.

The SFCP permits revising of the surveillance frequency for instrumentation channels. The NRC staff evaluated whether NEI 04-10 can be applied to subsets in an instrument channel when the SFCP currently specifies a surveillance interval that is applied to the entire channel. The NRC staff notes that the current channel surveillance may be performed by means of any series of sequential, overlapping, or total channel steps. In practice, this means that a channel is divided into subsets and each subset is tested separately. Therefore, the current instrument channel testing is already composed of a sequence of individual tests.

The instrument function may be modeled in the PRA differently depending on the site and the function (e.g., channel may be modeled individually, subsets may be modeled, or the channel function may be modeled as a single entity). There are different steps through the evaluation methodology in NEI 04-10 that could be used based on the different PRA modeling approaches.

The appropriate modeling of these different approaches is included in the NRC staffs review of the PRA modeling during the review of the application to implement an SFCP that uses NEI 04-10.

The PRA in use at Browns Ferry is the same as that was used to support the application that implemented an SFCP that uses NEI 04-10. The amendments will change the capability of the licensee to change the surveillance frequency of an entire channel to now change the frequency of each subset of the channel. The NRC staff finds that changes to the surveillance frequency caused by defining and using individual, testable component subsets can be appropriately evaluated with the current SFCP and the current PRAs. The NRC staff finds that the risk-informed methodology review and the PRA acceptability review that were performed during the review of the licensees application to implement an SFCP that uses NEI 04-10 is adequate and still applicable.

The NRC staff determined that the proposed changes to the TSs meet the requirements for TSs in 10 CFR 50.36(b). The regulations at 10 CFR 50.36 require that TSs include items in specified categories, including SRs. The proposed changes would modify the definitions applicable to instrumentation channel components but would not alter the technical approach that was approved by the NRC in NEI 04-10, and the TSs, as revised, continue to specify the appropriate SRs for tests and inspections to ensure the necessary quality of affected SSCs is maintained.

Additionally, the NRC staff finds the proposed TS changes to be technically clear and consistent with customary terminology and format in accordance with SRP Chapter 16.0. The NRC staff reviewed the proposed changes against the regulations and concludes that the changes continue to meet the requirements of Sections 50.36(b), 50.36(c)(3), and 50.36(c)(5) of 10 CFR, for the reasons discussed above, and thus provide reasonable assurance that the revised TSs provide the requisite requirements and controls for the facility to operate safely. Therefore, the NRC staff concludes that the proposed TS changes are acceptable.

3.3 Rescission of SR Consolidations The licensee, in previously approved amendments, consolidated the SRs for the performance of CHANNEL FUNCTIONAL TEST, CHANNEL CHECK, CHANNEL CALIBRATION, and LOGIC SYSTEM FUNCTIONAL TEST, because they had the common Frequency In accordance with the Surveillance Frequency Control Program, and appeared identical on the TS pages. These SRs are listed in section 2.3 of this safety evaluation.

The licensee stated in the LAR that, the magnitude of Browns Ferry documents that would require revision was not recognized and these consolidations would produce an error-likely situation because instrument functions with different frequencies were tracked under the same SR number within the SFCP. Therefore, the licensee proposed to restore the cited SR consolidations to their prior status, with a Frequency of, In accordance with the Surveillance Frequency Control Program.

The NRC staff reviewed the proposed SRs to be restored to their prior status and find them acceptable because the restoration of the associated SR frequencies is consistent with the intent of TSTF-425, Revision 3, the NRCs model safety evaluation dated July 6, 2009 (74 FR 31996), including the scope exclusions identified in Section 1.0, Introduction of the model safety evaluation. Also, restoration of the SRs for each instrument function will continue to meet the requirements of 10 CFR 50.36(c). The TS SRs proposed for inclusion in the SFCP either correspond to an equivalent SR in NUREG-1433 or are otherwise appropriated for inclusion in the program. The subject TS-specific SRs involve fixed periodic frequencies, and that in accordance with TSTF 425, changes to the frequencies for these SRs would be controlled under the SFCP.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Alabama State official was notified of the proposed issuance of the amendments on August 18, 2022. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The NRC has previously issued a proposed finding that the amendments involve no significant hazards consideration in the Federal Register on February 22, 2022 (87 FR 9653), and there has been no public comment on such finding.

Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

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

Principal Contributor: T. Sweat, NRR Date: November 21, 2022

ML22220A260 OFFICE NRR/DORL/LPL2-2/PM NRR/DORL/LPL2-2/LA NRR/DSS/STSB/BC NRR/DEX/EICB/BC NAME KGreen RButler VCusumano MWaters (RStattel for)

DATE 8/8/22 8/15/22 8/3/22 8/18/22 OFFICE OGC - NLO NRR/DORL/LPL2-2/BC NRR/DORL/LPL2-2/PM NAME JEzell DWrona KGreen DATE 8/26/22 11/21/22 11/21/22

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