ML100621404

Issuance of Amendments Regarding Deletion of Containment Isolation Valve Local Leak Rate Testing

ML100621404
Person / Time
Site:	Browns Ferry
Issue date:	03/22/2010
From:	Stewart Bailey Plant Licensing Branch II
To:	Krich R Tennessee Valley Authority
Bailey S , NRR/DORL,415-1321
References
TAC ME1801, TAC ME1802, TAC ME1803
Download: ML100621404 (21)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 March 22, 2010 Mr. R. M. Krich Vice President, Nuclear Licensing 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 DELETION OF CONTAINMENT ISOLATION VALVE LOCAL LEAK RATE TESTING (TAC NOS. ME1801, ME1802, AND ME1803)

Dear Mr. Krich:

The Commission has issued the enclosed Amendment Nos. 277, 304, and 263 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 July 27,2009. These amendments change Technical Specification Section 3.6.1.3, "Primary Containment Isolation Valves," to eliminate unnecessary local leak rate tests.

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

Sincerely,

~d Stewart N. Bailey, 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. 277 to DPR-33

2. Amendment No. 304 to DPR-52

3. Amendment No. 263 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. 277 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 July 27, 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. 277, 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 Douglas A. Broaddus, Acting 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: March 22, 2010

ATTACHMENT TO LICENSE AMENDMENT NO. 277 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 page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

REMOVE INSERT 3.6-16 3.6-16

-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. 277, 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-UI\IIT 1 Renewed License No. DPR-33 Amendment 277

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.5 Verify the isolation time of each power In accordance operated, automatic PCIV, except for MSIVs, with the Inservice is within limits. Testing Program SR 3.6.1.3.6 Verify the isolation time of each MSIV is 2': 3 In accordance seconds and :S: 5 seconds. with the Inservice Testing Program SR 3.6.1.3.7 Verify each automatic PCIV actuates to the 24 months isolation position on an actual or simulated isolation signal.

SR 3.6.1.3.8 Verify a representative sample of reactor 24 months instrumentation line EFCVs actuate to the isolation position on a simulated instrument line break signal.

SR 3.6.1.3.9 Remove and test the explosive squib from 24 months on a each shear isolation valve of the TIP System. STAGGERED TEST BASIS SR 3.6.1.3.10 Verify leakage rate through each MSIV is In accordance

S: 100 scfh and that the combined leakage with the Primary rate for all four main steam lines is :S: 150 scfh Containment when tested at 2': 25 psig. Leakage Rate Testing Program BFN-UNIT 1 3.6-16 Amendment Nos. 264 261, 277

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. 304 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 July 27, 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. 304, 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 Douglas A. Broaddus, Acting 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: March 22, 2010

ATTACHMENT TO LICENSE AMENDMENT NO. 304 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 page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

REMOVE INSERT 3.6-16 3.6-16

-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. 304, 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 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. DPR-52 Amendment 304

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.5 Verify the isolation time of each power In accordance operated, automatic PC IV, except for MSIVs, with the Inservice is within limits. Testing Program SR 3.6.1.3.6 Verify the isolation time of each MSIV is;::: 3 In accordance seconds and::;; 5 seconds. with the Inservice Testing Program SR 3.6.1.3.7 Verify each automatic PCIV actuates to the 24 months isolation position on an actual or simulated isolation signal.

SR 3.6.1.3.8 Verify a representative sample of reactor 24 months instrumentation line EFCVs actuate to the isolation position on a simulated instrument line break signal.

SR 3.6.1.3.9 Remove and test the explosive squib from 24 months on a each shear isolation valve of the TIP System. STAGGERED TEST BASIS SR 3.6.1.3.10 Verify leakage rate through each MSIV is In accordance

100 scfh and that the combined leakage with the Primary rate for all four main steam lines is

;; 150 scfh Containment when tested at ;::: 25 psig. Leakage Rate Testing Program BFN-Uf\lIT 2 3.6-16 Amendment No. 263, 267, 268, 304

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. 263 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 July 27,2009, 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-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. 263, 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 Douglas A. Broaddus, Acting 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: March 22, 2010

ATTACHMENT TO LICENSE AMENDMENT NO. 263 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 3.6-16 3.6-16

-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. 263, 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.

BFN-UI'JIT 3 Renewed License No. DPR-68 Amendment No. 263

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.5 Verify the isolation time of each power In accordance operated, automatic PCIV, except for MSIVs, with the Inservice is within limits. Testing Program SR 3.6.1.3.6 Verify the isolation time of each MSIV is ~ 3 In accordance seconds and::; 5 seconds. with the Inservice Testing Program SR 3.6.1.3.7 Verify each automatic PCIV actuates to the 24 months isolation position on an actual or simulated isolation signal.

SR 3.6.1.3.8 Verify a representative sample of reactor 24 months instrumentation line EFCVs actuate to the isolation position on a simulated instrument line break signal.

SR 3.6.1.3.9 Remove and test the explosive squib from 24 months on a each shear isolation valve of the TIP System. STAGGERED TEST BASIS SR 3.6.1.3.10 Verify leakage rate through each MSIV is In accordance

100 scfh and that the combined leakage with the Primary rate for all four main steam lines is

:; 150 scfh Containment when tested at ~ 25 psig. Leakage Rate Testing Program BFN-UNIT 3 3.6-16 Amendment No. 223, 227, 228, 263

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 277 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-33 AMENDMENT NO. 304 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-52 AMENDMENT NO. 263 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 July 27,2009, 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 would delete Surveillance Requirement (SR) 3.6.1.3.11 of TS 3.6.1.6, "Primary Containment Isolation Valves (PCIVs)." This SR requires verification that the combined leakage through water-tested lines that penetrate primary containment are within the limits specified in the BFN Primary Containment Leakage Rate Testing Program.

The licensee has been performing local leak rate tests (LLRTs) with water on a number of PCIVs that are in lines that either (1) terminate below the minimum post-loss of coolant accident (LOCA) suppression pool level, or (2) are in closed loops or closed systems outside primary containment.

TVA proposed to delete SR 3.6.1.3.11 and eliminate these tests on the basis that Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," and BFN TS 5.5.12, "Primary Containment Leakage Rate Testing Program," do not require these tests to be performed. Not performing these tests will avoid the associated personnel radiation dose.

2.0 REGULATORY EVALUATION

Section 50.36(c)(3) of 10 CFR Part 50 states that the TSs shall contain SRs 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, 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

-2 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 preparation against the draft of the 70 criteria current at the time of operating license application.

Section 50.54(0) to 10 CFR Part 50 requires the primary containment to be subject to the requirements of 10 CFR Part 50, Appendix J. Appendix J to 10 CFR Part 50 contains the requirements for leakage testing of the primary containment Option B of 10 CFR Part 50, Appendix J, identifies the performance-based requirements and criteria for leakage-rate testing, and specifies three types of tests; Type A (integrated leakage), Type B (penetration local leakage),

and Type C (PCIV local leakage). Appendix J defines a containment isolation valve (or PCIV) as "any valve which is relied upon to perform a containment isolation function." Containment is defined as "... an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment" Therefore, the staff considers an Appendix J PCIV to be a valve that could represent a potential fission product release pathway to the environment following a postulated accident BFI\J TS 5.5.12 requires TVA to maintain a primary containment leakage rate testing program in accordance with the guidelines contained in Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995. RG 1.163 endorses Nuclear Energy Institute (NEI) 94-01, Revision 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J," dated July 26, 1995, as providing methods acceptable to the NRC staff for complying with the provisions of 10 CFR Part 50, Appendix J, Option B. NEI 94-01 states that a LLRT is not required for the following cases:

• Primary containment boundaries that do not constitute potential primary containment atmospheric pathways during and following a Design-Basis Accident;

• Boundaries sealed with a qualified seal system; or,

• Test connection vents and drains between primary containment isolation valves which are one inch or less in size, administratively secured closed and consist of a double barrier.

NEI 94-01 also states that primary containment barriers that are sealed with a qualified seal system shall be periodically tested to demonstrate their functionality in accordance with the plant TSs. Leakage from containment isolation valves that are sealed with a qualified seal system may be excluded when determining the combined leakage rate provided that:

• Such valves have been demonstrated to have fluid leakage rates that do not exceed those specified in the TSs or associated bases, and

• The installed isolation valve seal-water system fluid inventory is sufficient to assume the sealing function for at least 30 days at a pressure of 1.1 times the LOCA peak accident pressure (1:1 Pa).

3.0 TECHNICAL EVALUATION

The qualified seal systems referred to in NEI 94-01 are systems that provide water fill either between containment isolation valves or between the disks in a containment isolation valve such that a water seal was assured for at least 30 days after system actuation. This water seal would preclude any containment atmosphere from escaping through that penetration line. The qualified

-3 seal systems typically have a relatively small delivery capacity, and the seal system pressure boundary leakage has to be correspondingly small to ensure the 30-day capability. In contrast, the suppression pool has a large volume of water available to prevent leakage of containment atmosphere from penetration lines that open below the suppression pool water level.

As described in NEI 94-01, LLRT is not required for penetrations that are not potential leakage pathways for primary containment atmosphere following the LOCA. Lines penetrating containment that open to the suppression pool below the minimum post-LOCA water level do not require LLRT because the suppression pool water acts as a continuous passive barrier (seal) preventing leakage of containment atmosphere. Lines analyzed as closed loops (systems) outside containment also do not need LLRT because they do not present a potential leakage pathway for the containment atmosphere.

In the application, TVA lists 38 locations where LLRTs would no longer be performed because they meet the above criteria for not being a potential leakage pathway for containment atmosphere. The locations are in the post accident sampling system, reactor core isolation cooling system, high pressure coolant injection system, residual heat removal system, and core spray system. TVA stated that these locations have historically been tested to the requirements of a qualified seal system.

The PCIVs in the closed systems that also function as pressure isolation valves will continue to be tested as required by BFN TS 5.5.6, "Inservice Testing Program," which includes valve exercise, position indication verification, and leak rate testing, as applicable. Closed systems or loops outside containment will continue to be subject to BFN TS 5.5.2, "Primary Coolant Sources Outside Containment," and be monitored for pressure boundary leakage by periodic visual inspection and system leakage tests.

Therefore, the staff finds that the licensee's proposed deletion of TS SR 3.6.1.3.11 to be acceptable based on continued compliance with existing regulations and the BFN license requirements regarding containment leakage rate testing.

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 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 changes 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 Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (74 FR 53781, October 20,2009). 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.

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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 Contributor: Jerome Bettie Date: l"1arch 22, 2010

ML100621404 NRR-058 OFFICE LPL2-2/PM LPL2-2/LA ITSB/BC SCVB/BC OGC NLO LPL2-2/BC (A) w/comments NAME SBailey BClayton RElliott RDennig* AJones DBroaddus DATE 03/17/10 03/08/10 3/2/10 2/14/10 03/16/10 03/22/10

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