ML14051A726
| ML14051A726 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 02/28/2014 |
| From: | Farideh Saba Plant Licensing Branch II |
| To: | James Shea Tennessee Valley Authority |
| Saba F NRR/DORL 301-415-1447 | |
| References | |
| TAC MF3124, TAC MF3125, TAC MF3126 | |
| Download: ML14051A726 (3) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Joseph W. Shea Vice President-Nuclear Licensing Tennessee Valley Authority 1101 Market Street, LP 3D-C Chattanooga, TN 37402-2801 February 28, 2014
SUBJECT:
BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2, AND 3 - REQUEST FOR ADDITIONAL INFORAMTION REGARDING PROPOSED TECHNICAL SPECIFICATION CHANGE TO REVISE THE LEAKAGE RATE THROUGH MAIN STEAM ISOLATION VALVES-TS-485 (TAC NOS. MF3124, MF3125, AND MF3126)
Dear Mr. Shea:
By letter dated November 22, 2013, Tennessee Valley Authority (TVA, the licensee) requested amendments to Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 for the Browns Ferry Nuclear Plant Units 1, 2, and 3 respectively. The licensee proposed a technical specification change to revise the leakage rate through the main steam isolation valves.
A response to the enclosed request for additional information (RAI) is needed before the Nuclear Regulatory Commission (NRC) staff can complete the review. This request was discussed with Mr. William Gordon of your staff on February 7, 2014, and it was agreed that TVA would respond by March 21, 2014. The NRC staff considers that timely responses to RAis help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources.
If you have any questions, please contact me by phone at (301) 415-1447 or e-mail at Farideh.Saba@ nrc.gov.
Docket Nos. 50-259, 50-260, and 50-296
Enclosure:
Request for Additional Information cc: Distribution via Listserv Sincerely, Farideh E. Saba, Senior Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
REQUEST FOR ADDITIONAL INFORMATION (RAI)
PROPOSED TECHNICAL SPECIFICATION CHANGE TO REVISE THE LEAKAGE RATE THROUGH MAIN STEAM ISOLATION VALVES-TS-485 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2, AND 3 DOCKET NOS. 50-259, 50-260, AND 50-296 By letter dated November 22, 2013 {Agencywide Documents Access Management System Accession No. ML14015A403), Tennessee Valley Authority requested amendments to Renewed Facility Operating License Numbers DPR-33, DPR-52, and DPR-68 for the Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, respectively. The licensee proposed a technical specification change to revise the leakage rate through the main steam isolation valves.
Background
The primary and secondary Alternate Leakage Treatment pathway share a common passive valve, FCV-1-57, which is normally open with power removed. This represents a single point failure for both systems.
BFN Units 1, 2, and 3 lnservice Testing (1ST) Program is currently operating in its fourth interval with the Code of Record of American Society of Mechanical Engineers, Operating and Maintenance Code, 2004 Edition through 2006 Addenda. The BFN 1ST program has determined that the FCV-1-57 valve is a passive, Category B type valve due to the fact that it is normally open with power removed during normal operation.
The FCV-1-57 is a motor-operated valve that fails as is. The 1ST program requirement for a passive Category B valve is position indication verification per section ISTC-3700, which states "Valves with remote position indicators shall be observed locally at least once every 2 years to verify that valve operation is accurately indicated. Where practicable, this local observation should be supplemented by other indications, such as use of flowmeters or other suitable instrumentation to verify obturator position. These observations need not be concurrent. Where local observation is not possible, other indications shall be used for valve operation."
- 1) Please explain how BFN is meeting the requirement of ISTC-3700 for valve FCV-1-57.
- 2) Please explain how valve stem and disc integrity of valve FCV-1-57 is being monitored and maintained.
Enclosure
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