Text

Issuance of Amendments 274, 301 and 260 Regarding Control Rod Notch Test Frequency (TS-462)
	ML091050704
Person / Time
Site:	Browns Ferry
Issue date:	05/29/2009
From:	Ellen Brown; Plant Licensing Branch II
To:	Swafford P; Tennessee Valley Authority
	Brown E, NRR/DORL, 415-2315
References
	TAC ME0024, TAC ME0025, TAC ME0026
	Download: ML091050704 (20)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 29,2009 Mr. Preston D. Swafford Chief Nuclear Officer and Executive Vice President 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801

SUBJECT:

BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2, AND 3 - ISSUANCE OF AMENDMENTS REGARDING CONTROL ROD NOTCH TEST FREQUENCY (TAC NOS. ME0024, ME0025, AND ME0026) (TS-462)

Dear Mr. Swafford:

The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment Nos. 274, 301 and 260 to Renewed Facility Operating Licenses 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 October 30, 2008, as supplemented by a letter dated November 20, 2008. The amendments adopt the applicable portions of Technical Specification Task Force Traveler TSTF-475, Revision 1, Control Rod Notch Testing Frequency and SRM

[source range monitor] Insert Control Rod Action, which was approved for use in a safety evaluation dated November 5, 2007. More specifically, the amendments revise the Technical Specification (TS) Surveillance Requirement (SR) 'frequency for notch testing of each fully withdrawn control rod from weekly to monthly, as well as clarify in a TS example that the 1.25 surveillance test interval extension in SR 3.0.2 is also applicable to time periods discussed in SR Notes.

However, should the SR frequency relaxation result in a noticeable trend in failures, the licensee is expected to consider the need for revising the TS to include a more conservative testing frequency in accordance with NRC Administrative Letter 98-10, Dispositioning of Technical Specifications that are Insufficient to Assure Plant Safety.

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. 274 to DPR 33

2. Amendment No. 301 to DPR 52

3. Amendment No. 260 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.274 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 October 30,2008, as supplemented by letter dated November 20,2008, 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. 274, 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 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Thomas H. Boyce, hief 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: May 29, 2009

ATTACHMENT TO LICENSE AMENDMENT NO. 274 TO 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.

Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

REMOVE INSERT 1.4-5 1.4-5 3.1-8 3.1-8 3.1-10 3.1-10

-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. 274, 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.

BFN-UNIT 1 Renewed License No. DPR-33 Amendment No. 274

Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-3 (continued)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

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

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after 2: 25% RTP, Perform channel adjustment. 7 days The interval continues whether or not the unit operation is

< 25% RTP between performances, As the Note modifies the required performance of the Surveillance, it is construed to be part of the "specified Frequency," Should the 7 day interval be exceeded while operation is < 25% RTP, this Note allows 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after power reaches 2: 25% RTP to perform the Surveillance, The Surveillance is still considered to be within the "specified Frequency." Therefore, if the Surveillance were not performed within the 7 day (plus the extension allowed by SR 3.0.2) interval, but operation was < 25% RTP, it would not constitute a failure of the SR or failure to meet the LCO, Also, no violation of SR 3,0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (plus the extension allowed by SR 3.0.2) with power 2: 25% RTP.

Once the unit reaches 25% RTP, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months would be allowed for completing the Surveillance. If the Surveillance were not performed within this 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval (plus the extension allowed by SR 3.0.2), there would then be a failure to perform a Surveillance within the specified Frequency and the provisions of SR 3.0.3 would apply.

(continued)

BFN-UNIT 1 1.4-5 Amendment No. 2-64; 209, 274

Control Rod OPERABILITY 3.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A (continued) A.3 Perform SR 3.1.3.3 for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from each withdrawn discovery of OPERABLE control rod. Condition A concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM AND A4 Perform SR 3.1.1 .1. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

8. Two or more withd rawn B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months control rods stuck.

C. One or more control rods C.1 -------------~OTE------------

inoperable for reasons RWM may be bypassed other than Condition A or as allowed by

8. LCO 3.3.2.1, if required, to allow insertion of inoperable control rod and continued operation.

Fully insert inoperable 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months controI rod.

Mill C.2 Disarm the associated 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months CRD.

(continued)

BFN-UNIT 1 3.1-8 Amendment No. ~, 274

Control Rod OPERABILITY 3.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.3.1 Determine the position of each control rod. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months SR 3.1.3.2 (Deleted).

SR 3.1.3.3 --------------------------N0TE------------------------

Not required to be performed until 31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM.

Insert each withdrawn control rod at least one 31 days notch.

SR 3.1.3.4 Verify each control rod scram time from fully In accordance withdrawn to notch position 06 is s 7 with SR 3.1.4.1, seconds. SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 (continued)

BFN-UNIT 1 3.1-10 Amendment No. ~, 274

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. 301 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 October 30, 2008, as supplemented by letter dated November 20, 2008, 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 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-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. 301, 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 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

~

Thomas H. Boyce, Chi4/

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: ~1ay 29, 2009

ATTACHMENT TO LICENSE AMENDMENT NO. 301 TO 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.

Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

REMOVE INSERT 1.4-5 1.4-5 3.1-8 3.1-8 3.1-10 3.1-10

-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. 301, 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 this 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. 301

Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-3 (continued)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

~----------------------NOTE-------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after?:: 25% RTP.

Perform channel adjustment 7 days The interval continues whether or not the unit operation is

< 25% RTP between performances.

As the Note modifies the required performance of the Surveillance, it is construed to be part of the Il specified Frequency." Should the 7 day interval be exceeded while operation is < 25% RTP, this Note allows 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after power reaches ~ 25% RTP to perform the Surveillance. The Surveillance is still considered to be within the "specified Frequency." Therefore, if the Surveillance were not performed within the 7 day (plus the extension allowed by SR 3.0.2) interval, but operation was < 25% RTP, it would not constitute a failure of the SR or failure to meet the LeO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (plus the extension allowed by SR 3.0.2) with power 225% RTP.

Once the unit reaches 25% RTP, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months would be allowed for completing the Surveillance. If the Surveillance were not performed within this 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval (plus the extension allowed by SR 3.0.2), there would then be a failure to perform a Surveillance within the specified Frequency and the provisions of SR 3.0.3 would apply.

(continued)

BFN-UNIT 2 1.4-5 Amendment No. ~ r66, 301

Control Rod OPERABILITY 3.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A (continued) A3 Perform SR3.1.3.3 for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from each withdrawn discovery of OPERABLE control rod. Condition A concurrent with THERMAL POWER greater than the low power setpo int (LPSP) of the RWM AND A.4 Perform SR 3.1.1.1. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months B. Two or more withdrawn B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months control rods stUCk.

C. One or more control rods C.1 ~------------NOTE------------

inoperable for reasons RWM may be bypassed other than Condition A or as aI/owed by B. LCO 3.3.2.1, if required, to allow insertion of inoperable control rod and continued operation.

Fully insert inoperable 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months control rod.

AND C.2 Disarm the associated 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months CRD.

(continued)

BFN-UNIT 2 3.1-8 Amendment No. 259,301

Control Rod OPERABILITY 3.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Determine the position of each control rod. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months SR 3.1.3.2 (Deleted).

SR 3.1.3.3 ~--------~---------~------NOTE-------------------------

Not required to be performed until 31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM.

Insert each withdrawn control rod at least one 31 days notch.

SR 3.1.3.4 Verify each control rod scram time from fully In accordance withdrawn to notch position 06 is s 7 with SR 3.1.4.1 ,

seconds. SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 (continued)

BFN-UNIT 2 3.1-10 Amendment No. rse, 301

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. 260 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 October 30,2008, as supplemented by letter dated November 20,2008, 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. Accordillgly, 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. 260, 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 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

~~

Thomas H. Boyce, Chiei)

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: r~ay 29, 2009

\

)

ATIACHMENT TO LICENSE AMENDMENT NO. 260 TO 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.

Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

REMOVE INSERT 1.4-5 1.4-5 3.1-8 3.1-8 3.1-10 3.1-10

-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. 260, 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 tt:le 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.

BFN-UNIT 3 Renewed License No. DPR-68 Amendment No. 260

Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-3 (continued)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

N()TE-------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after? 25% RTP.

Perform channel adjustment. 7 days The interval continues whether or not the unit operation is

< 25% RTP between performances.

As the Note modifies the required performance of the Surveillance, it is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is < 25% RTP, this Note allows 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after power reaches? 25% RTP to perform the Surveillance. The Surveillance is still considered to be within the "specified Frequency." Therefore, if the Surveillance were not performed within the 7 day (plus the extension allowed by SR 3.0.2) interval, but operation was < 25% RTP, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (plus the extension allowed by SR 3.0.2) with power

~25% RTP.

()nce the unit reaches 25% RTP, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months would be allowed for completing the Surveillance. If the Surveillance were not performed within this 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval (plus the extension allowed by SR 3.0.2), there would then be a failure to perform a Surveillance within the specified Frequency and the provisions of SR 3.0.3 would apply.

(continued)

BFN-UNIT 3 1.4-5 Amendment No..~~, 260

Control Rod OPERABiliTY 3.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A3 Perform SR 3.1.3.3 for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from each withdrawn discovery of OPERABLE control rod. Condition A concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM AND AA Perform SR 3.1.1.1. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months B. Two or more withdrawn B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months control rods stUCk.

C. One or more control rods C.1 ------------NOTE-----------

inoperable for reasons RWM may be bypassed other than Condition A or as allowed by B. LCO 3.3.2.1, if required, to allow insertion of inoperable control rod and continued operation.

Fully insert inoperable 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months control rod.

AND C.2 Disarm the associated 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months CRD.

(continued)

BFN-UNIT 3 3.1-8 Amendment No. e4r, 260

Control Rod OPERABILITY 3.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.3.1 Determine the position of each control rod. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months SR 3.1.3.2 (Deleted).

SR 3.1.3.3 --------------------------N()TE------------------------

Not required to be performed until 31 days after the control rod is withdrawn and THERMAL P()WER is greater than the LPSP of the RWM.

Insert each withdrawn control rod at least one 31 days notch.

SR 3.1.3.4 Verify each control rod scram time from fully In accordance withdrawn to notch position 06 is .~ 7 with SR 3.1.4.1, seconds. SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 (continued)

BFN-UNIT 3 3.1-10 Amendment No. 2-+2-,260

\."R REGU( UNITED STATES

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~ 0.,. NUCLEAR REGULATORY COMMISSION

,:; .L WASHINGTON, D.C. 20555-0001 t-

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- - - i' SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 274 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-33 AMENDMENT NO. 301 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-52 AMENDMENT NO. 260 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, 50-296

1.0 INTRODUCTION

By letter dated October 30, 2008 as supplemented by a letter dated November 20, 2008, the Tennessee Valley Authority (TVA, the licensee) submitted a license amendment request regarding Browns Ferry Nuclear (BFN) Plant, Units 1, 2, and 3. The proposed amendments adopt the applicable portions of Technical Specification Task Force Traveler (TSTF)-475, Revision 1, Control Rod Notch Testing Frequency and SRM Insert Control Rod Action, which was approved for use in a safety evaluation dated November 5,2007. More specifically, the amendments revise the Technical Specification (TS) Surveillance Requirement (SR) frequency for notch testing of each fully withdrawn control rod from weekly to monthly, as well as clarify in a TS example that the 1.25 surveillance test interval extension in SR 3.0.2 is also applicable to time periods discussed in SR Notes.

The licensee's supplementary submittal dated November 20, 2008, provided clarifying information did not change the scope of the October 3D, 2008, proposed amendment. The supplemental information was included in the original notice of proposed action published in the Federal Register on February 24, 2009.

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 the 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.

On November 13,2007, the U.S. Nuclear Regulatory Commission (NRC, the Commission) announced the availability of TSTF-475 in the Federal Register (72 FR 63935). The TSTF involves three changes to the Standard TS NUREGs that, depending upon the adopting plant, mayor may not be adopted by a plant. The first changes the surveillance frequency for control

-2 rod notch testing from 7 to 31 days. The second adds the word "fully" to a REQUIRED ACTION statement to clarify that control rods should be fully inserted, and applies to only BWR

[boiling-water reactor]/6 plants (the BFN units are BWR/4s). The third change clarifies the usage of the 1.25 surveillance frequency interval extension.

Appendix A to 10 CFR 50, General Design Criterion (GDC) 29, Protection against anticipated occurrence, requires that the protection and reactivity control systems be designed to assure an extremely high probability of accomplishing their safety functions in an event of anticipated operational occurrences. The design relies on the control rod drive system (CRDS) to function in conjunction with the protection systems under anticipated operational occurrences, including loss of power to all recirculation pumps, tripping of the turbine generator, isolation of the main condenser, and loss of all offsite power. The CRDS provides an adequate means of inserting sufficient negative reactivity to shut down the reactor and prevent exceeding acceptable fuel design limits during anticipated operational occurrences. Meeting the requirements of GDC 29 for the CRDS prevents occurrence of mechanisms that could result in fuel cladding damage such as severe overheating, excessive cladding strain, or exceeding the thermal margin limits during anticipated operational occurrences. Preventing excessive cladding damage in the event of anticipated transients ensures maintenance of the integrity of the cladding as a fission product barrier.

However, the BFN units were designed and constructed based on the proposed GDC published by the Atomic Energy Commission in the Federal Register (32 FR 10213) on July 11, 1967 (draft GDC). TVA reviewed the differences between the draft GDC and final GDC. As discussed in the NRC Staff Requirements Memorandum for SECY-92-223, the Commission decided not to apply the final GDC to plants with construction permits issued prior to May 21, 1971. As the BFN units were licensed before the final GDC were formally adopted, these units were evaluated on a plant-specific basis, determined to be safe, and licensed by the Commission. The draft GDC applicable to these units is maintained in Appendix A to the Updated Final Analyses Report (UFSAR). In this case draft GDC-29 generally encompasses the design requirements for final GDC 29.

NRC Administrative Letter (AL) 98-10, Dispositioning of Technical Specifications that are Insufficient to Assure Plant Safety, states that "the discovery of an improper or inadequate TS value or required action is considered a degraded or nonconforming condition," which is defined in NRC Inspection Manual Chapter 9900; see latest guidance in Regulatory Issue Summary 2005-20, Revision to NRC Inspection Manual Part 9900 Technical Guidance, "Operability Determinations & Functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality or Safety." AL 98-10 also states "Imposing administrative controls in response to an improper or inadequate TS is considered an acceptable short-term corrective action. The NRC staff expects that, following the imposition of administrative controls, an amendment to the inadequate TS, with appropriate justification and schedule, will be submitted in a timely fashion."

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3.0 TECHNICAL EVALUATION

3.1 Revising the SR Frequency for Notch Testing of Each Fully Withdrawn Control Rod from Weekly to Monthly The CROS is the primary reactivity control system for the reactor. The CROS, in conjunction with the Reactor Protection System, provides the means for the reliable control of reactivity changes to ensure under all conditions of normal operation, including anticipated operational occurrences that specified acceptable fuel design limits are not exceeded. Control rods are components of the CROS that have the capability to hold the reactor core subcritical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the CROS.

The CROS consists of a Control Rod Orive Mechanism (CROM), by which the control rods are moved, and a hydraulic control unit (HCU) for each control rod. The CROM is a mechanical hydraulic latching cylinder that positions the control blades. The collet piston mechanism design feature ensures that the control rod will not be inadvertently withdrawn. This is accomplished by engaging the collet fingers, mounted on the collet piston, in notches located on the index tube.

Oue to the tapered design of the index tube notches, the collet piston mechanism will not impede rod insertion under normal insertion or scram conditions.

The collet retainer tube (CRT) is a short tube welded to the upper end of the CRO that houses the collet mechanism, which consist of the locking collet, collet piston, collet return spring and an unlocking cam. The collet mechanism provides the locking/unlocking mechanism that allows the insert/withdraw movement of the control rod. The CRT has three primary functions: (a) to carry the hydraulic unlocking pressure to the collet piston, (b) to provide an outer cylinder, with a suitable wear surface for the metal collet piston rings, and (c) to provide mechanical support for the guide cap, a component that incorporates the cam surface for holding the collet fingers open and also provides the upper rod guide or bushing.

CRT cracking was first discovered in 1975. It was determined that during scrams, the CRT temperature distribution changes substantially at reactor operating conditions. Relatively cold water moves upward through the inside of the CRT and exits via the flow holes into the annulus on the outside. At the same time hot water from the reactor vessel flows downward on the outside surface of the CRT. There is very little mixing of the cold water flowing from the three flow holes into the annulus and the hot water flowing downward. Thus, there are substantial through wall and circumferential temperature gradients during scrams, which contribute to the observed CRT cracking. It was recognized that notch testing provided a method to demonstrate the integrity of the CRT. Each partially or fully withdrawn operable control rod was required to be exercised one notch at least once each week. Control rod insertion capability was demonstrated by inserting each partially or fUlly withdrawn control rod at least one notch and observing that the control rod moves. The control rod may then be returned to its original position. This ensures the control rod is not stuck and is free to insert on a scram signal.

Subsequently, many BWRs have reduced the frequency of notch testing for partially withdrawn control rods from weekly to monthly. The notch test frequency for fully withdrawn control rods was still performed weekly. The change, for partially withdrawn control rods, was made because of the potential power reduction required to allow control rod movement for partially

- 4 withdrawn control rods, the desire to coordinate scheduling with other plant activities, and the fact that a large sample of control rods are still notch tested on the weekly basis. The operating experience related to the changes in CRD performance also provided additional justification to reduce the notch test frequency for the partially withdrawn control rods. Current operating experience now provides justification to reduce the notch test frequency for the fully withdrawn control rods as well. A review of industry operating experience did not identify any incidents of stuck control rods identified via performance of rod notch surveillance for either partially or fully withdrawn control rods. Therefore, increasing the CRD notch testing frequency for fully withdrawn control rods from weekly to monthly will have minimal impact on the reliability of the CRD system.

Although not a basis for approving the frequency extension of notch testing for partially withdrawn control rods, General Electric (GE) Nuclear Energy report, "CRD Notching Surveillance Testing for Limerick Generating Station," provides additional insight as to why a review of industry operating experience may not have identified any incidents of stuck control rods identified via performance of rod notch surveillance. The GE report is discussed in TSTF-475, Revision 1. The GE report provides a description of the cracks noted on the original design CRT surfaces. These cracks, which were later determined to be intergranular, were generally circumferential, and appeared with greatest frequency below and between the cooling water ports, in the area of the change in wall thickness. Subsequently, cracks associated with residual stresses were also observed in the vicinity of the attachment weld. Continued circumferential cracking could lead to 360 degree severance of the CRT that would render the CRD inoperable, which would prevent insertion, withdrawal or scram. Such failure would be detectable in any fully or partially withdrawn control rod during the surveillance notch testing required by the TSs. To a lesser degree, cracks have also been noted at the welded joint of the interim design CRT but no cracks have been observed in the final improved CRT design. No collet housing failures have been noted since 1975. In addition, the intergranular stress corrosion cracking (IGSCC) growth rates were evaluated using GE's PLEDGE model with the assumption that the water chemistry condition is based on GE recommendations. The model is based on fundamental principles of stress corrosion cracking, which can evaluate crack growth rates as a function of water oxygen level, conductivity, material sensitization and applied loads.

It was determined that the additional time of 24 days represented an additional 10 mils of growth in total crack length. The small difference in growth rate would have little effect on the behavior between one notch test and the next subsequent test. Therefore, from the materials perspective based on low crack growth rates, a decrease in the notch test frequency would not affect the reliability of detecting a CRDM failure due to crack growth.

In addition to notch testing, other SRs are performed to verify the operability of the CRDS.

Scram time testing can identify failure of individual CRD operation resulting from IGSCC initiated cracks and mechanical binding. Unlike the CRD notch tests, these single rod scram tests cover the other mechanical components such as scram pilot solenoid operated valves, the scram inlet and outlet air operated valves, and the scram accumulator, as well as operation of the control rods. Thus, the primary assurance of scram system reliability is provided by the scram time testing since it monitors the system scram operation and the complete travel of the control rod. Also, the HCUs, CRD drives, and control rods are also tested during refueling outages, approximately every 18-24 months. Based on the data collected during the preceding cycle of operation, selected CRDs are inspected and, as required, their internal components are replaced. As a result, increasing the CRD notch testing frequency of fully withdrawn rods from

-5 weekly to monthly will have minimal impact on the reliability of the CRDS since additional SRs are performed that verify the operability of the system.

Based on no known CRD failures having been detected during the notch testing SR, as well as the performance of other diverse SRs used to verify the operability of the CRDS, the NRC staff concludes that revising the SR frequency for notch testing of each fully withdrawn control rod from weekly to monthly is acceptable.

It should be noted that approval to relax the SR frequency for notch testing of each fully withdrawn control rod is based on, in part, operational experience that has demonstrated no known CRD failures having been detected during the notch testing SR. Should the SR frequency relaxation result in a noticeable trend in failures, the licensee is expected to consider the need for revising the TS to include a more conservative testing frequency in accordance with NRC AL 98-10. AL 98-10 states that "Occasionally, as a result of licensee design-basis reconstitution efforts or NRC inspection efforts, licensees determine that specific values or required actions in TS may not assure safety. When this occurs, licensees typically conduct an evaluation and, if necessary, institute administrative controls that instruct the operators to maintain a more restrictive value for a particular parameter or to take a more conservative required action." AL 98-10 also goes on to state that "Imposing administrative controls in response to an improper or inadequate TS is considered an acceptable short-term corrective action. The NRC staff expects that, following the imposition of administrative controls, an amendment to the TS, with appropriate justification and schedule, will be submitted in a timely fashion."

3.2 Clarify in TS Example that the 1.25 Surveillance Test Interval Extension in SR 3.0.2 is also Applicable to Time Periods Discussed in SR Notes The NRC staff has reviewed the proposal to amend Example 1.4-3 in Section 1.4 "Frequency,"

to clarify that the 1.25 provision in SR 3.0.2 is equally applicable to time periods specified in the Notes of the "Surveillance" column. The NRC staff finds this change acceptable since the revision clarifies the example to make it consistent with the definition of specified "Frequency" provided in the second paragraph of Section 1.4, which states that "the 'specified Frequency' is referred to throughout this section and each of the Specifications of Section 3.0, Surveillance Requirement (SR) Applicability. The 'specified Frequency' consists of the requirements of the Frequency column of each SR, as well as certain Notes in the Surveillance columns that modify performance requirements."

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 and changes SRs.

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,

-6 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 8288). 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. However, should the SR frequency relaxation result in a noticeable trend in failures, the licensee is expected to consider the need for revising the TS to include a more conservative testing frequency in accordance with NRC AL 98-10.

Principal Contributor: A. Lewin Date: May 29, 2009

ML091050704 NRR-058 OGC NLO OFFICE LPL2-2/PM LPL-WB/LA ITSB/BC LPL2-2/BC w/ comment NAME EBrown BClayton RElliott

AJones TBoyce DATE 04/28/09 04/27/09 02/05/09 05/13/09 OS/29/09

By Memo

v t e Letter Sequence Approval
TAC:ME0024, Control Rod Notch Testing Frequency and SRM Insert Control Rod Action (Approved, Closed) TAC:ME0025, Control Rod Notch Testing Frequency and SRM Insert Control Rod Action (Approved, Closed) TAC:ME0026, Control Rod Notch Testing Frequency and SRM Insert Control Rod Action (Approved, Closed)
Initiation Acceptance... Administration Questions Answers Results Approval Other: ML091760260
MONTHYEAR2009-05-29 [Table View]

ML091050704

Text

SUBJECT:

Dear Mr. Swafford:

Enclosures:

Attachment:

Attachment:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

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