ML091620129

Watts Bar Nuclear Plant, Unit 1, License Amendment, Changed Technical Specification 3.5.3,Containment Isolation Valves, Regarding Position Verification Requirements (TAC No. ME1435)

ML091620129
Person / Time
Site:	Watts Bar
Issue date:	09/03/2009
From:	Lamb J G Watts Bar Special Projects Branch
To:	Swafford P D Tennessee Valley Authority
Lamb J G, NRR/DORL, 415-1727
References
TAC ME1435
Download: ML091620129 (16)

Text

UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 September 3, 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 WATTS BAR NUCLEAR PLANT, UNIT 1 -ISSUANCE OF AMENDMENT REGARDING THE APPLICATION FOR CHANGE TO TECHNICAL SPECIFICATION 3.6.3, "CONTAINMENT ISOLATION VALVES," REGARDING POSITION VERIFICATION REQUIREMENTS (TAC NO. ME1435)

Dear Mr. Swafford:

The Commission has issued the enclosed Amendment No. 79 to Facility Operating License No. NPF-90 for Watts Bar Nuclear Plant, Unit 1. This amendment is in response to your application dated June 5, 2009, as supplemented July 10, 2009. The proposed amendment would provide alternatives for valve position verification in various Required Actions and Surveillance Requirements in TS 3.6.3, "Containment Isolation Valves." The proposed TS changes are based on TS Task Force (TSTF) change traveler TSTF-269-A (Revision 2, approved July 27, 1999) and TSTF-45-A (Revision 2, approved July 26, 1999). The purpose of TSTF-269-A, Revision 2 was to modify requirements for repetitive verification of the status of "locked, sealed, or secured" components to allow the verification to be "by administrative means." The purpose of TSTF-45-A, Revision 2, was to modify TS 3.6 valve line up surveillance requirements to specify that only containment isolation valves which are not locked, sealed, or otherwise secured are required to verified closed. TSTF-269-A, Revision 2, and TSTF-45-A, Revision 2, were incorporated into NUREG-1431, "Standard Technical Specifications:

Westinghouse Plants," published June 2004. A copy of the safety evaluation is also enclosed.

Notice of issuance will be included in the Commission's biweekly Federal Register notice. Si.ncerel..yf/J/? f Z1/]

zV-J G. Lamb, Manager . s Bar Special Projects Branch D ision of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-390

Enclosures:

1. Amendment No. 79 to NPF-90 2. Safety Evaluation ccw/encls:

Distribution via Listserv UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-390 WATTS BAR NUCLEAR PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 79 License No. NPF-90 1. The Nuclear Regulatory Commission (the Commission) has found that: The application for amendment by the Tennessee Valley Authority (the licensee) dated June 5, 2009, as supplemented July 10, 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 Title 10 Code of Federal Regulations (10 CFR) Chapter I; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 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; 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 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.

-2Accordingly, 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 Facility Operating License No. NPF-90 is hereby amended to read as follows: Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 79 ,and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. This license amendment is effective as of the date of its issuance, and shall be implemented no later than 60 days from the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Watts Bar Special Projects Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Changes to License No. NPF-90 the Technical Date of Issuance:

September 3, 2009 ATTACHMENT TO LICENSE AMENDMENT NO. FACILITY OPERATING LICENSE NO. DOCKET NO. Replace page 3 of Operating License No. NPF-90 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 area of change. REMOVE INSERT 3.6-9 3.6-9 3.6-10 3.6-10 3.6-11 3.6-11 3.6-12 3.6-12

-3TVA, 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, instrument calibration, or other activity associated with radioactive apparatus or components; and TVA, pursuant to the Act and 10 CFR Parts 30,40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified or incorporated below. Maximum Power Level TVA is authorized to operate the facility at reactor core power levels not in excess of 3459 megawatts thermal. Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 79 ,and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. Safety Parameter Display System (SPDS) (Section 18.2 of SER Supplements 5 and 15) Prior to startup following the first refueling outage, TVA shall accomplish the necessary activities, provide acceptable responses, and implement all proposed corrective actions related to having the Watts Bar Unit 1 SPDS operational. Vehicle Bomb Control Program (Section 13.6.9 of SSER 20) During the period of the exemption granted in paragraph 2.D.(3) of this license, in implementing the power ascension phase of the approved initial test program, TVA shall not exceed 50% power until the requirements of 10 CFR 73.55(c)(7) and (8) are fully implemented.

TVA shall submit a letter under oath or affirmation when the requirements of 73.55(c)(7) and (8) have been fully implemented.

Amendment No. 79 Containment Isolation Valves 3.6.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued)

A.2

1. Isolation devices in high radiation areas may be verified by use of administrative means. 2. Isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative means.

Verify the affected penetration flow path is isolated.

Once per 31 days for isolation devices outside containment AND Prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days for isolation devices inside containment B. -------------N 0 TE Only applicable to penetration flow paths with two containment isolation valves.

One or more penetration flow paths with two containment isolation valves inoperable except for purge valve or shield building bypass leakage not within limit. B.1 Isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (continued)

Watts Bar-Unit 1 3.6-9 Amendment J..9 Containment Isolation Valves 3.6.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. -------------N 0 TE Only applicable to penetration flow paths with only one containment isolation valve and a closed system.

One or more penetration flow paths with one containment isolation valve inoperable.

C.1 AND C.2 Isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange.

1. Isolation devices in high radiation areas may be verified by use of administrative means. 2. Isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative means.

Verify the affected penetration flow path is isolated.

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Once per 31 days D. Shield building bypass not within limit. 0.1 Restore leakage within limit. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> E. One or more penetration flow paths with one or more containment purge valves not within purge valve leakage limits. E.1 AND Isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (continued)

Watts Bar-Unit 1 3.6-10 Amendment Containment Isolation Valves 3.6.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. (continued)

E.2 AND E.3 ---------------

NOT E

1. Isolation devices in high radiation areas may be verified by use of administrative means. 2. Isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative means. ----------------------

..

Verify the affected penetration flow path is isolated.

Perform SR 3.6.3.5 for the resilient seal purge valves closed to comply with Required Action E.1. Once per 31 days for isolation devices outside containment AND Prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days for isolation devices inside containment Once per 92 days Once per 92 days F. Required Action and associated Completion Time not met. F.1 AND Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> F.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Watts Bar-Unit 1 3.6-11

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.1 Verify each containment purge valve is closed, except when the containment purge valves are open for pressure control, ALARA or air quality considerations for personnel entry, or for Surveillances that require the valves to be open. 31 days SR 3.6.3.2 Valves and blind flanges in high radiation areas may be verified by use of administrative controls.

Verify each containment isolation manual valve and blind flange that is located outside containment, the containment annulus, and the Main Steam Valve Vault Rooms, and not locked, sealed or otherwise secured and required to be closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls.

31 days SR 3.6.3.3 Valves and blind flanges in high radiation areas may be verified by use of administrative means. Verify each containment isolation manual valve and blind flange that is located inside containment, the containment annulus, and the Main Steam Valve Vault Rooms, and not locked, sealed or otherwise secured and required to be closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls.

Prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days (continued)

Watts Bar-Unit 1 3.6-12 Amendment .J3 UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REGULATION RELATED TO AMENDMENT NO. 79 TO FACILITY OPERATING LICENSE NO. NPF-90 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT, UNIT 1 DOCKET NO. 50-390

1.0 INTRODUCTION

By letter dated June 5, 2008 (Agencywide Document and Management System No. ML091610068), as supplemented July 10,2009 (ML091940188), the Tennessee Authority (TVA, licensee) submitted a request for changes to the Technical Specifications for Watts Bar Nuclear Plant (WBN), Unit The proposed amendment would provide alternatives for valve position verification in Required Actions and Surveillance Requirements in TS 3.6.3, "Containment Isolation The proposed TS changes are based on TS Task Force (TSTF) change traveler (Revision 2, approved July 27, 1999) and TSTF-45-A (Revision 2, approved July The purpose of TSTF-269-A, Revision 2 was to modify requirements for repetitive verification the status of "locked, sealed, or secured" components to allow the verification to be administrative means." The purpose of TSTF-45-A, Revision 2, was to modify TS 3.6 valve up surveillance requirements (SRs) to specify that only containment isolation valves (CIVs) are not locked, sealed, or otherwise secured are required to be verified closed.

Revision 2, and TSTF-45-A, Revision 2, were incorporated into NUREG-1431, Technical Specifications:

Westinghouse Plants," published June The licensee's supplementary submittal dated July 10, 2009, provided clarifying information did not change the scope of the proposed amendment as described in the original notice proposed action published in the Federal Register on June 30,2009 (74 FR 31327) and did change the initial proposed no significant hazards

2.0 REGULATORY EVALUATION

The provisions of Title 10, Code of Federal Regulations (10 CFR) Section 50.36, "Technical Specifications," include requirements for the contents of TSs. These shall include SRs relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

-2 WBN Unit 1 was designed to meet the intent of the "Proposed General Design Criteria (GDC) for Nuclear Power Plant Construction Permits" published in July, 1967. The WBN construction permit was issued in January 1973. WBN Unit 1 Updated Final Safety Analysis Report (UFSAR) addresses 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," as published in July 1971, including Criterion 4 as amended October 27, 1987. Criterion 16 -Containment Design Reactor containment and associated systems shall be provided to establish an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require. Criterion 50 -Containment Design Basis The reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and, with sufficient margin, the calculated pressure and temperature conditions resulting from any LOCA [Ioss-of-coolant accident].

This margin shall reflect consideration of (1) the effects of potential energy sources which have not been included in the determination of the peak conditions, such as energy in steam generators and energy from metal-water and other chemical reactions that may result from degraded emergency core cooling functioning, (2) the limited experience and experimental data available for defining accident phenomena and containment responses, and (3) the conservatism of the calculational model and input parameters.

Criterion 54 -Piping Systems Penetrating Containment Piping systems penetrating primary reactor containment shall be provided with leak detection, isolation, and containment capabilities having redundance, reliability, and performance capabilities which reflect the importance to safety of isolating these piping systems. Such piping systems shall be designed with a capability to test periodically the operability of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits. Criterion 55 -Reactor Coolant Pressure Boundary Penetrating Containment Each line that is part of the reactor coolant pressure boundary and that penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis: One locked closed isolation valve inside and one locked closed isolation valve outside containment; or One automatic isolation valve inside and one locked closed isolation valve outside containment; or One locked closed isolation valve inside and one automatic isolation valve outside containment.

A simple check valve may not be used as the automatic isolation valve outside containment; or

-One automatic isolation valve inside and one automatic isolation valve outside containment.

A simple check valve may not be used as the automatic isolation valve outside containment.

Criterion 56-Primary Containment Isolation Each line that connects directly to the containment atmosphere and penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis: One locked closed isolation valve inside and one locked closed isolation valve outside containment; or One automatic isolation valve inside and one locked closed isolation valve outside containment; or One locked closed isolation valve inside and one automatic isolation valve outside containment.

A simple check valve may not be used as the automatic isolation valve outside containment; or One automatic isolation valve inside and one automatic isolation valve outside containment.

A simple check valve may not be used as the automatic isolation valve outside containment.

Criterion 57 -Closed Systems Isolation Valves Each line that penetrates primary reactor containment and is neither part of the reactor coolant pressure boundary nor connected directly to the containment atmosphere shall have at least one CIV which shall be either automatic, or locked closed, or capable of remote manual operation.

This valve shall be outside containment and located as close to the containment as practical.

A simple check valve may not be used as the automatic isolation valve. Exceptions to GOC 54, 55, 56, and 57 The UFSAR states that requirements and intent of U.S. Nuclear Regulatory Commission (NRC) GOC 54, 55, 56. and 57 have been met with the following exceptions: Primary containment monitoring instrument systems shall be designed to maintain the integrity of the containment isolation boundary in the event of a design-basis event (OBE). The reactor vessel level indication system (RVLlS) is required post accident for continual indication of the water level in the reactor vessel. The reactor coolant system (RCS) wide range pressure transmitter (PT-68-70) is required post accident for continual indication of the pressure in the reactor vessel. Containment isolation for each RHR sump line penetration consists of: A closed system outside containment. A containment isolation valve outside containment in each of the two lines after the penetrating line branches in the RHR sump valve room. Both of these valves are remotely controlled from the main control room.

-The pressure boundary valve leak rate test line CIVs (63-158,63-112,63-111,63-167,63-174, 63-21, and 63-121) are remote manually actuated from the main control room and do not receive a containment isolation signal. The design configuration for penetrations X-79A (ice blowing) and X-79B (negative return) is temporarily modified in operating Modes 5 and 6 and when the reactor is defueled (Mode 7) to support ice blowing activities.

The TS changes contained in TVA's license amendment request, dated June 5, 2009, were evaluated by comparing them to the applicable criteria as stated above. The amendment was also compared to the Standard TSs (STSs) in NUREG-1431 for consistency.

The exceptions stated above are immaterial to the license amendment request, dated June 5, 2009, as supplemented July 10, 2009. TVA did not propose a change to the physical design or operation of any containment systems in its license amendment request, dated June 5, 2009, as supplemented July 10, 2009. TECHNICAL EVALUATION The NRC staff previously reviewed TSTF-45, Revision 2, and TSTF-269, Revision 2. TSTF-45, Revision 2, and TSTF-269, Revision 2 were approved by the NRC staff on July 26, 1999 and July 27, 1999, respectively.

The CIVs form part of the containment pressure boundary and provide a means for fluid penetrations not serving accident consequence limiting systems to be provided with two isolation barriers that are closed on a containment isolation signal. These isolation devices are either passive or active (automatic).

Manual valves, deactivated automatic valves secured in their closed position (including check valves with flow through the valve secured), blind flanges, and closed systems are considered passive devices. Check valves, or other automatic valves designed to close without operator action following an accident, are considered active devices. Two barriers in series are provided for each penetration so that no single credible failure or malfunction of an active component can result in a loss of isolation or leakage that exceeds limits assumed in the safety analyses.

One of these barriers may be a closed system. These barriers (typically CIVs) make up the Containment Isolation System. The operability requirements for TS 3.6.3 regarding CIVs help ensure that containment is isolated within the time limits assumed in the safety analyses.

Therefore, the operability requirements of TS 3.6.3 provide assurance that the containment function assumed in the safety analyses will be maintained.

TVA proposed the follOWing changes to Required Actions A.2, C.2, and E.2 in TS 3.6.3: Add the number 1 to the existing note in the text box, Add the number 2 to the new note in the text box, Change "NOTE" to "NOTES," Add a note that states, "Isolation devices that are locked, sealed, or secured may be verified by use of administrative

-5 TSTF-269-A, Revision 2, states the following:

It is sufficient to assume that the initial establishment of component status (e.g., isolation valves closed) was performed correctly.

Subsequently verification is intended to ensure the component has not been inadvertently repositioned.

Given that the function of locking, sealing or securing components is to ensure the same avoidance of inadvertent repositioning, the periodic re-verification should only be a verification of the administrative control that ensures that the component remains in the required state. It would be inappropriate to remove the lock, seal, or other means of securing the component solely to perform an active verification of the required state. Therefore, the NRC staff finds that allowing the position of locked, sealed, or otherwise secured valves to be verified by administrative means, is acceptable.

SR 3.6.3.2 requires verification that each containment isolation manual valve and blind flange located outside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed. The SR helps to ensure that post-accident leakage of radioactive fluids or gases outside of the containment boundary are within design limits. This SR does not require any testing or valve manipulation.

Rather, it involves verification that those CIVs outside containment and capable of being mispositioned are in the correct position.

The SR specifies that CIVs that are open under administrative controls are not required to meet the SR during the time the valves are open. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing.

SR 3.6.3.3 requires verification that each containment isolation manual valve and blind flange located inside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed. The SR helps to ensure that post-accident leakage of radioactive fluids or gases outside of the containment boundary are within design limits. The SR specifies that CIVs that are open under administrative controls are not required to meet the SR during the time they are open. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing.

TVA proposed to add the following words to Surveillance Requirement SR 3.6.3.2 and SR 3.6.3.3: * "and not locked, sealed, or otherwise secured." The NRC staff finds excluding locked, sealed, or otherwise secured valves from periodic position verification is acceptable because administrative controls exist for such valves. The NRC staff finds that these controls have proven to be adequate to ensure that the valves are maintained in the positions required by the plant safety analyses when primary and secondary containment are required to be operable.

-6 The proposed TS 3.6.3 changes to the Required Action A.2, Required Action Co2, and Required Action E.2, SR 3.6.3.2, and SR 3.6.3.3 for WBN Unit 1 continue to ensure that the containment is isolated within the time limits assumed in the safety analyses and the respective Limiting Condition of Operations are met. TVA proposed no changes to the physical design or operation of containment systems. Therefore, the NRC staff finds that the regulatory requirements described in the Regulatory Evaluation section above continue to be met with the proposed changes. The NRC staff finds that the proposed changes are also consistent with the STSs. The NRC staff finds the proposed changes acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Tennessee 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, 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 June 30,2009 (74 FR 31327). 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. Principle Matt Hamm John G. Lamb Date: September 3, 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 WATTS BAR NUCLEAR PLANT, UNIT 1 -ISSUANCE OF AMENDMENT REGARDING THE APPLICATION FOR CHANGE TO TECHNICAL SPECIFICATION 3.6.3, "CONTAINMENT ISOLATION VALVES," REGARDING POSITION VERIFICATION REQUIREMENTS (TAC NO. ME1435)

Dear Mr. Swafford:

The Commission has issued the enclosed Amendment No. 79 to Facility Operating License No. NPF-90 for Watts Bar Nuclear Plant, Unit 1. This amendment is in response to your application dated June 5, 2009, as supplemented July 10, 2009. The proposed amendment would provide alternatives for valve position verification in various Required Actions and Surveillance Requirements in TS 3.6.3, "Containment Isolation Valves." The proposed TS changes are based on TS Task Force (TSTF) change traveler TSTF-269-A (Revision 2, approved JUly 27, 1999) and TSTF-45-A (Revision 2, approved July 26, 1999). The purpose of TSTF-269-A, Revision 2 was to modify requirements for repetitive verification of the status of "locked, sealed, or secured" components to allow the verification to be "by administrative means." The purpose of TSTF-45-A, Revision 2, was to modify TS 3.6 valve line up surveillance requirements to specify that only containment isolation valves which are not locked, sealed, or otherwise secured are required to verified closed. TSTF-269-A, Revision 2, and TSTF-45-A, Revision 2, were incorporated into NUREG-1431, "Standard Technical Specifications:

Westinghouse Plants," published June 2004. A copy of the safety evaluation is also enclosed.

Notice of issuance will be included in the Commission's biweekly Federal Register notice. Sincerely, IRAI John G. Lamb, Senior Project Manager Watts Bar Special Projects Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-390

Enclosures:

1. Amendment No. 79 to NPF-90 2. Safety Evaluation ccw/encls:

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