ML19351E870
| ML19351E870 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 11/13/1980 |
| From: | Mills L TENNESSEE VALLEY AUTHORITY |
| To: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| Shared Package | |
| ML19351E866 | List: |
| References | |
| NUDOCS 8012190472 | |
| Download: ML19351E870 (3) | |
Text
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II 400 Chestnut Street 'Itwer II November 13, 1980 Mr. Jame's P. O'Reilly, Director Office od Inspection and Ehforcement U.S. Nuclear Regulatory Otanission Region II - Suite 3100 101 Marietta Street Atlanta, Georgia 30303
Dear Mr. O'Reilly:
Enclosed is our response to your Cetober 27, 1980, letter to H. G. Parris RII: PAT 50-259/80-28, -260/80-21, and -296/80-22 concerning activities at Browns Ferry Nuclear Plant which appeared to be in rchpliance with NIC requirements. 'Ihis supplements the response subnitted by my letter to you dated September 4, 1980.
We have reviewed the above inspection report ant ilnd no proprietary ir.fornation in it.
If you have any questions, please call Jim Daner at EP3 857-2014.
Wry truly yours, TDeIESSE VAIL 5% AUWORI'IY L. M. Mills, Manager Nuclear Regulation and Safety Enclosure 8012190
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EPCIOSURE RP"IONSE 'IO J. P. O'REILLY'S IEITER DATED OCTOBER 27,1980, 'IO H. G. PARRIS BROWNS FERRY NUCLEAR PIN.T (RII:PA" 50-259/80-28, 50-260/80-21, 50-296/80-22)
Appendix A of the inspection letter identified activities which were apparently in ncoccupliance with NBC requirements. We following are those items and our response.
Infraction As required by 10 CFR 50 Appendix B and implemented by
'Dennessee Valley Authority Nuclear - Operational Quality Assurance Manual Section 2.1 Part II, upon ccupletion of maintenance on any item of the CSSC list and before release for service, appropriate testing shall be performed to verify operational acceptability.
Contrary to the above, on June 23, 1980, the control valve number 2 pressure sensor, which serds a signal to the Reactor Protection System to indicate turbine control valve fast closure, had maintenance performed on it and was returned to -
service without testing of the Reactor Protection System to
- verify proper operation of the repaired pressure switch.
~ Response Cerrective Steps 'Paken and Results Achieved te subject pressure switch, located in the reactor protection system, is autcmatically bypassed and not in service below 30-percent power.
(
Reference:
Browns Ferry 'Itchnical Specification 3.1-A, Table 3.1.A, " Operability of
'Ibrbine Control Valve Ioss of Oil Pressure Trip Etnction at Power Levels 30 percent.)" Bypass of this scram function below 30-percent power is provided by a pressure switch that monitors turbine first-stage pressure.
Testing of the subject pressure switch is normally performed above 30-percent l
Power by procedure S. I. 4.1.A.12.
Testing of the subject reactor protection system pressure switch was acccuplished as soon as possible after the 30 percent automatic bypass interlock was cleared.
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Corrective Steps Taken to Avoid Further Monempliance Events '. elated to the failure of the subject pressure switch (PS-47-144) have been reported under Licensee Event Report 50-259/8050. She event has been discussed with employees involved regarding failure to identify the subject pressure switch as safety-related (CSSC) equipnent. Affected employees have been properly cautioned regarding the need to accurately identify all safety-related maintenanc ~.~rk.
A copy of the CSSC listing fr m the OQN4 has been placed in the shift engineers office to assist in identifying CSSC equipnent a.,d to ensure that prescribed maintenance methods and post-mainter. w e testing are used where necessary. Additionally, the TVA Division of tt.: lear Power Quality Assurance Staff has investigated the incident. The results of this investigation are now being evaluated for future corrective action. This evaluation should be emplete by Jutuary 1, 1981.
Date B.211 Cttpliance Was Achieved Full empliance was achieved on Cx:tober 14, 1980.
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