ML092370045

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Issuance of Amendments Regarding Technical Specification Improvement to Adopt Technical Specification Task Force (TSTF) TSTF-476, Revision 1
ML092370045
Person / Time
Site: Browns Ferry  Tennessee Valley Authority icon.png
Issue date: 11/19/2009
From: Ellen Brown
Plant Licensing Branch II
To: Swafford P
Tennessee Valley Authority
Brown E, NRR/DORL, 415-2315
References
TAC ME1214, TAC ME1215, TAC ME1216, TS-464
Download: ML092370045 (22)
v • d • e


Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 November 19, 2009 Mr. Preston D. Swafford Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 3R Lookout Place 1101 Market Street Chattanooga, TN 37402-2801

SUBJECT:

BROWNS FERRY NUCLEAR PLANT, UNITS 1,2, AND 3 -ISSUANCE OF AMENDMENTS REGARDING TECHNICAL SPECIFICATION IMPROVEMENT TO ADOPT TECHNICAL SPECIFICATION TASK FORCE (TSTF) TSTF-476, REVISION 1, REGARDING BANKED POSITION WITHDRAWAL SEQUENCE CONTROL ROD INSERTION PROCESS (TAC NOS. ME1214, ME1215, AND ME1216) (TS-464)

Dear Mr. Swafford:

The Commission has issued the enclosed Amendment Nos. 276, 303, and 262 to Renewed Facility Operating Licenses Nos. DPR-33, DPR-52, and DPR-68 for the Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, respectively. These amendments are in response to your application dated April 24, 2009. The changes in these amendments are related Technical Specifications (TSs) Sections 3.1.6, Rod Pattern Control, and 3.3.2.1, Control Rod Block Instrumentation to allow the BFN units to reference the improved control rod banked position withdrawal sequence (BPWS) when performing a reactor shutdown. In addition, the changes would add a footnote to TS Table 3.3.2.1-1, Control Rod Block Instrumentation.

The changes are consistent with Nuclear Regulatory Commission approved Industry Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-476, Revision 1, Improved BPWS Control Rod Insertion Process (NEDO-33091). The availability of this TS improvement was announced in the Federal Register on May 23, 2007 (72 FR 29004) as part of the consolidated line item improvement process.

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

Sincerely, IRA!

Eva A. Brown, Senior Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-259, 50-260, and 50-296

Enclosures:

1. Amendment No. 276 to DPR-33
2. Amendment No. 303 to DPR-52
3. Amendment No. 262 to DPR-68
4. Safety Evaluation cc w/enclosures: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 276 Renewed License No. DPR-33

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

A. The application for amendment by Tennessee Valley Authority (the licensee) dated April 24, 2009, 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.

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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 as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 276, are hereby incorporated in the 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 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

--:l ._y /" Thomas H. Boyc ,Chief r Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Operating License and Technical Specifications Date of Issuance: November 19, 2009

ATTACHMENT TO LICENSE AMENDMENT NO. 276 RENEWED FACILITY OPERATING LICENSE NO. DPR-33 DOCKET NO. 50-259 Replace Page 3 of Renewed Operating License DPR-33 with the attached Page 3.

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 3.3-20 3.3-20

-3 (3) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts 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 3458 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendices A and S, as revised through Amendment No. 276, 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.

SFN-UNIT 1 Renewed License No. DPR-33 Amendment No. 276

Control Rod Block Instrumentation 3.3.2.1 Table 3.3.2.1-1 (page 1 of 1)

Control Rod Block Instrumentation APPLICABLE MODES OR FUNCTION OTHER REQUIRED SURVEILLANCE ALLOWABLE SPECIFIED CHANNELS REQUIREMENTS VALUE CONDITIONS

1. Rod Block Monitor
a. Low Power Range - Upscale (a) 2 SR 3.3.2.1.1 (e)

SR 3.3.2.1.4 SR 3.3.2.1.8

b. Intermediate Power Range - Upscale (b) 2 SR 3.3.2.1.1 (e)

SR 3.3.2.1.4 SR 3.3.2.1.8

c. High Power Range - Upscale (I),(g) 2 SR 3.3.2.1.1 (e)

SR 3.3.2.1.4 SR 3.3.2.1.8

d. Inop (g).(h) 2 SR 3.3.2.1.1 .NA
e. Downscale (g).(h) 2 SR 3.3.2.1.1 (i)

SR 3.3.2.1.4

2. Rod Worth Minimizer 1(c),2(c) SR 3.3.2.1.2 NA SR 3.3.2.1.3 SR 3.3.2.1.5 SR 3.3.2.1.7
3. Reactor Mode Switch Shutdown Position (d) 2 SR 3.3.2.1.6 NA (a) THERMAL POWER ~ 27% and:'> 62% RTP and MCPR less than the value specified in the COLR.

(b) THERMAL POWER> 62% and :s 82% RTP and MCPR less than the value specified in the COLR.

(c) With THERMAL POWER 510% RTP. except during the reactor shutdown process if the coupling of each withdrawn control rod has been confirmed.

(d) Reactor mode switch in the shutdown position.

(e) Less than or equal to the Allowable Value specified in the COLR.

(f) THERMAL POWER> 82"10 and < 90% RTP and MCPR less than the value specified In'the COLR.

(g) THERMAL POWER ~ 90% RTP and MCPR less than the value specified in the COLR.

(h) THERMAL POWER ~ 27% and < 90% RTP and MCPR less than the value specified in the COLR.

(I) Greater than or equal to the Allowable Value specified in the COLR.

BFN-UNIT 1 3.3-20 Amendment No. 2d4, 262, 276

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 303 Renewed License No. DPR-52

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

A. The application for amendment by Tennessee Valley Authority (the licensee) dated April 24, 2009, 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.

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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 as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 303 ,are hereby incorporated in the 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 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

rhomas~ L AA!.I~

1 'i'I Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Operating License And Technical Specifications Date of Issuance: November 19, 2009

ATTACHMENT TO LICENSE AMENDMENT NO. 303 RENEWED FACILITY OPERATING LICENSE NO. DPR-52 DOCKET NO. 50-260 Replace Page 3 of Renewed Operating License DPR-52 with the attached Page 3.

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 3.3-21 3.3-21

-3 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 3458 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 303, 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 OPR-52, the first performance is due at the end of the first surveillance interval that begins at implementation of 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 t~lis amendment shall include the relocation of these requirements to the appropriate documents, as described in the licensee's BFN-UNIT 2 Renewed License No. OPR-52 Amendment No. 303

ContrOl Rod Block Instrumentation 3.3.2.1 Table 3.3.2.1-1 (page 1 of 1)

Control Rod Block Instrumentation APPLICABLE MODES OR FUNCTION OTHER REQUIRED SURVEILLANCE ALLOWABLE SPECIFIED CHANNELS REQUIREMENTS VALUE CONDITIONS

1. Rod Block Monitor
a. Low Power Range - Upscale (a) 2 SR 3.3.2.1.1 (e)

SR 3.3.2.1.4 SR 3.3.2.1.8

b. Intennediate Power Range* Upscale (b) 2 SR 3.3.2.1.1 (e)

SR 3.3.2.1.4 SR 3.3.2.1.8

c. High Power Range* Upscale (f), (g) 2 SR 3.3.2.1.1 (e)

SR 3.3.2.1.4 SR 3.3.2.1.8

d. Inop (g),{h) 2 SR 3.3.2.1.1 NA
e. Downscale (g),(h) 2 SR 3.3.2.1.1 (I)

SR 3.3.2.1.4

2. Rod Worth Minimizer 1(C),2{C) SR 3.3.2.1.2 NA SR 3.3.2.1.3 SR 3.3.2.1.5 SR 3.3.2.1.7
3. Reactor Mode SwItch
  • Shutdown Position (d) 2 SR 3.3.2.1.6 NA (a) THERMAL POWER 2: 27% and ~ 62% RTP and MCPR less than the value speclflEld in the COLR .

(b) THERMAL POWER> 62% and ~ 82% RTP and MCPR less than the value specified in'the COLR .

(c) With THERMAL POWER ~ 10% RTP, except during the reactor shutdown process If the coupling of each withdrawn control rod has been confinned.

(d) Reactor mode switch In the shutdown position.

(e) Less than or equal to the Allowable Value specified in the COLR.

(f) THERMAL POWER> 82% and < 90% RTP and MCPR less than the value specified In the COLR.

(g) THERMAL POWER 2: 90% RTP and MCPR less than the value specified in the COLR.

(h) THERMAL POWER 2: 27% and < 90% RTP and MCPR less than the value specified iii the COLR.

(I) Greater than or equal to the Allowable Value specifted In the COLR.

BFN-UNIT 2 3.3-21 Amendment No. ~, 303

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 262 Renewed License No. DPR-68

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

A. The application for amendment by Tennessee Valley Authority (the licensee) dated April 24, 2009, 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

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 as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 262, are hereby incorporated in the 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 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION L~H1+V.Ar-J

¥ Thomas H. yce, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Operating License And Technical Specifications Date of Issuance: November 19, 2009

ATTACHMENT TO LICENSE AMENDMENT NO. 262 RENEWED FACILITY OPERATING LICENSE NO. DPR-68 DOCKET NO. 50-296 Replace Page 3 of Renewed Operating License DPR-68 with the attached Page 3.

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 3.3-21 3.3-21

-3 (3) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts 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 3458 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 262, 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.

BFI\J-UNIT 3 Renewed License No. DPR-68 Amendment No. 262

Control Rod Block Instrumentation 3.3.2.1 Table 3.3.2.1-1 (page 1 of 1)

Control Rod Block Instrumentation APPLICABLE MODES OR FUNCTION OTHER REQUIRED SURVEILLANCE ALLOWABLE SPECIFIED CHANNELS REQUIREMENTS VALUE CONDITIONS

1. Rod Block Monitor
a. Low Power Range *,Upscale (a) 2 SR 3.3.2.1.1 (e)

SR 3.3.2.1.4 SR 3.3.2.1.8

b. Intermediate Power Range - Upscale (b) 2 SR 3.3.2.1.1 (e)

SR 3.3.2.1.4 SR 3.3.2.1.8

c. High Power Range - Upscale (f), (g) 2 SR 3.3.2.1.1 (e)

SR 3.3.2.1.4 SR 3.3.2.1.8

d. lnap (g),(h) 2 SR 3.3.2.1.1 NA
e. Downscale (g),(h) 2 SR 3.3.2.1.1 (I)

SR 3.3.2.1.4

2. Rod Worth Minimizer 1(c),2(c) SR 3.3.2.1.2 NA SR 3.3.2.1.3 SR 3.3.2.1.5 SR 3.3.2.1.7
3. Reactor Mode Switch - Shutdown Position (d) 2 SR 3.3.2.1.6 NA (a) THERMAL POWER ~ 27% and ~ 62% RTP and MCPR less than the value specified In'the COLR.

(b) THERMAL POWER> 62% and ~82% RTP and MCPR less than the value specified In*the COLR.

(c) With THERMAL POWER ~ 10% RTP, except during the reactor shutdown process If the coupling of each withdrawn control rod has been confirmed.

(d) Reactor mode switch in the shutdown position.

(e) Less than or equal to the Allowable Value specified in the COLR.

(f) THERMAL POWER> 82% and < 90% RTP and MCPR less than the value specified in the COLR.

(g) THERMAL POWER ~ 90% RTP and MCPR less than the value specified in the COLR.

(h) THERMAL POWER ~ 27% and < 90% RTP and MCPR less than the value specified in the COLR.

(i) Greater than or equal to the Allowable Value specified In the COLR.

BFN-UNIT 3 3.3-21 Amendment No. e+a, 262

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 SAFETY EVALUATION BYTHE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 276 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-33 AMENDMENT NO. 303 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-52 AMENDMENT NO. 262 TO RENEWED FACILITY OPERATING LICENSE NO. 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 letter dated April 24, 2009, Tennessee Valley Authority (TVA, the licensee) proposed changes to the technical specifications (TSs) for the Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3.

The changes are the adoption of TS Task Force (TSTF)-476, Revision 1, Improved [Banked Position Withdrawal Sequence) BPWS Control Rod Insertion Process (NEDO-33091-A). The TSTF allows use of the improved BPWS during shutdowns if the conditions of NEDO-33091-A, Revision 2,Improved BPWS Control Rod Insertion Process, dated July 2004, have been satisfied.

This TSTF involves changes to NUREG-1433, General Electric Plants, BWR [Boiling-Water Reactory4, Rev. 3 STS [Standard TS}, Vol. 1, Specifications, Section 3.1.6, Rod Pattern Control, Section 3.3.2.1, Control Rod Block Instrumentation, and Table 3.3.2.1-1. BFN's TS Tables 3.3.2.1 are identical to TS Table 3.3.2.1 in STSs NUREG-1433. The BFN units are all BWRJ4 plants.

2.0 REGULATORY EVALUATION

Section 50.36(c)(3) of Title 10 to the Code of Federal Regulations (10 CFR) states that the TS shall contain surveillances related to the test, calibration, or inspection to assure that necessary quality for systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

The construction permits for BFN Units 1, 2, and 3 predate the formal issuance of the current Appendix A, General Design Criteria (GDC), to 10 CFR Part 50. During the construction permit licensing process, Units 1 and 2 were evaluated against the Comment Draft of 27 Criteria, which was issued on November 22, 1965, while Unit 3 was evaluated against the Comment Draft of 70 Criteria, which was issued on July 10, 1967. The design bases of each BFN unit were reevaluated at the time of initial Final Safety Analysis Report (FSAR) preparation against the draft of the 70 criteria current at the time of operating license application.

-2 As stated in Appendix A of BFN Updated FSAR, based on the understanding of the intent of the proposed criteria current at the time of operating license application, BFN conforms with the intent of the Atomic Energy Commission GDC for Nuclear Power Plant Construction Permits. As the GDC were finalized, the requirements were placed in Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities. GDC 28 applies to the control rod insertion process. GDC 28 states that the reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects o"f postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures, or other reactor pressure vessel internals to impair significantly the capability to cool the core. The comparable criterion for the BFN units is draft GDC-32, which states that:

Limits, which include considerable margin, shall be placed on the maximum reactivity worth of control rods or elements and on rates at which reactivity can be increased to ensure that the potential effects of a sudden or large change ofreactivity cannot (a) rupture the reactor coolant pressure boundary or (b) disrupt the core, its support structures, or other vessel internals sufficiently to impair the effectiveness of emergency core cooling.

2.1 TSTF-476 This TSTF involves implementation of an improved BPWS, which allows licensees of General Electric BWRs to follow the improved BPWS when inserting control rods into the core during a reactor shutdown. By letter dated January 9,2007, the BWR Owners Group proposed these changes for incorporation into the STSs as TSTF-476. These changes are based on the Nuclear Regulatory Commission (NRC) staff-approved topical report NEDO-33091-A, Improved BPWS Control Rod Insertion Process, Revision 2 dated July 2004, as approved by NRC in a safety evaluation dated June 16, 2004.

The control rod drop accident (CRDA) is the design-basis accident for the subject TS changes. In order to minimize the impact of a CRDA, the BPWS process was developed to minimize control rod reactivity worth for BWR plants. The proposed improved BPWS further simplifies the control rod insertion process, and in order to evaluate it, the NRC staff followed the guidelines of NUREG-0800, Standard Review Plan for the Review of Safety Analyses Reports for Nuclear Power Plants, Section 15.4.9, and referred to GDC-28, which is comparable to draft-GDC-32, as its regulatory requirement.

2.2 Proposed TS Changes The proposed changes to the BFN TS Include:

  • Revising TS Table 3.3.2.1-1, "Control Rod Block Instrumentation, " to add a footnote that allows operators to bypass the rod worth minimizer if conditions for the optional BPWS shutdown process are satisfied. New footnote (c) states, "With THERMAL POWER::::; 10% RTP, except during the reactor shutdown process if the coupling of each withdrawn control rod has been confirmed."

-3

3.0 TECHNICAL EVALUATION

In its safety evaluation for Licensing Topical Report NEDO-33091-A, the NRC staff determined that the methodology described in TSTF-476, Revision 1, to incorporate the improved BPWS into the STSs, is acceptable.

TSTF-476, Revision 1 states that the improved BPWS provides the following benefits: (1) allows the plant to reach the all-rods-in condition prior to significant reactor cool down, which reduces the potential for re-criticality as the reactor cools down; (2) reduces the potential for an operator reactivity control error by reducing the total number of control rod manipulations; (3) minimizes the need for manual scrams during plant shutdowns, resulting in less wear on control rod drive (CRD) system components and CRD mechanisms; and, (4) eliminates unnecessary control rod manipulations at low power, resulting in less wear on reactor manual control and CRD system components.

The BFN units are BWR/4 with original/standard BPWS already implemented. Therefore, BFN meets the conditions in NEDO-33091-A to use the improved BPWS, and the potential for a CRDA with power below the low-power setpoint (LPSP) has been eliminated. The safety evaluation for NEDO-33091-A explained that the potential for the CRDA will be eliminated by the following changes to operational procedures, which BFN has committed to make prior to implementation:

1. Before reducing power to the LPSP, operators shall confirm control rod coupling integrity for all rods that are fully withdrawn. Control rods that have not been confirmed coupled and are in intermediate positions, must be fully inserted prior to power reduction to the LPSP. No action is required for fully-inserted control rods.

If a shutdown is required and all rods, which have not confirmed coupled, cannot be fully inserted prior to the power dropping below the LPSP, then the original/standard BPWS must be adhered to.

2. After reactor power drops below the LPSP, rods may be inserted from notch position 48 to notch position 00 without stopping at the intermediate positions. As stated in the NRC's safety evaluation of NEDO-33091-A, General Electric Nuclear Energy recommends that operators insert rods in the same order as specified for the original/standard BPWS as much as is reasonably possible. If a plant is in the process of shutting down following improved BPWS with the power below the LPSP, no control rod shall be withdrawn unless the control rod pattern is in compliance with standard BPWS requirements.

-4 In addition to the procedure changes specified above, the NRC staff previously verified during its review of NEDO-33091-A, Revision 2 that no single failure of the BWR CRD mechanical or hydraulic system can cause a control rod to drop completely out of the reactor core during the shutdown process. Therefore, the proper use of the improved BPWS will prevent a CRDA from occurring while power is below LPSP.

The NRC staff finds the proposed TS changes in TVA's amendment request properly incorporate the improved BPWS procedure into the BFN TS, and that BFN accurately adopted TSTF-476 and the requisite procedural changes. Therefore, the NRC staff approves TVA's license amendment request to adopt TSTF-476, Revision 1.

The licensee included in its application the revised TS Bases to be implemented with the TS change. The NRC notes that the TS Bases Control Program is the appropriate process for updating the affected TS Bases.

4.0 STATE CONSULTATION

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

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (74 FR 37249, July 28, 2009). Accordingly, the amendment meets 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 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) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: K. Bucholtz R. Grover Date: November 19, 2009

ML092370045 NRR-058 OFFICE LPL2-2/PM LPL2-2/LA ITSB/BC SRXB/BC OGC "NLO" LPL2-2/BC NAME EBrown BClayton RElliott* GCranston* JSuttenberg LRaghavan fnr ~R{ v{'p, DATE 10/13/09 11/19/09 8/12/09 8/12/09 10/14/09 11/19/09

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