05000260/LER-2003-004, Regarding Cable Separations Design Error Related to Appendix Requirements
| ML032480970 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 09/05/2003 |
| From: | Bhatnagar A Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 03-004-00 | |
| Download: ML032480970 (8) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 2602003004R00 - NRC Website | |
text
September 5, 2003 U.S. Nuclear Regulatory Commission 10 CFR 50.73 ATTN: Document Control Desk Mail Stop OWFN, P1-35 Washington, D. C. 20555-0001
Dear Sir:
TENNESSEE VALLEY AUTHORITY - BROWNS FERRY NUCLEAR PLANT (BFN) -
UNITS 2 and 3 - DOCKETS 50-260 AND -296 - FACILITY OPERATING LICENSES DPR - 52 AND DPR LICENSEE EVENT REPORT (LER) 50-260/2003-004-00 The enclosed report provides details of an electrical cable separations design error which potentially affected the plants Appendix R safe-shutdown capability.
In accordance with 10 CFR 50.73(a)(2)(ii)(B), TVA is reporting this event as an unanalyzed condition that could have significantly degraded plant safety.
There are no commitments contained in this letter.
Sincerely, Original signed by M. D. Skaggs,for:
Ashok S. Bhatnagar cc: See page 2
U.S. Nuclear Regulatory Commission Page 2 September 5, 2003 TEA:DTL:PSH:BAB Enclosure cc (Enclosure):
A. S. Bhatnagar, PAB 1B-BFN M. J. Burzynski, BR 4X-C M. D. Skaggs, POB 2C-BFN J. E. Maddox, LP 6A-C R. F. Marks, PAB 1C-BFN J. Scott Martin, PMB 1A-BFN F. C. Mashburn, BR 4X-C D. C. Olcsvary, LP 6A-C C. L. Root, PAB 1G-BFN K. W. Singer, LP 6A-C E. J. Vigluicci, ET 11A-K R. E. Wiggall, PEC 2A-BFN LEREvents@inpo.org NSRB Support, LP 5M-C EDMS-K S:lic/submit/lers/260-2003-04.doc
Abstract
Appendix R-related calculations associated with the restart of BFN Unit 1, it was noted that some associated circuits of certain 4 kV electrical distribution boards and loads were not adequately evaluated in the Unit 2 and Unit 3 calculations.
As physically configured, control power cables which affect power circuit breaker control are routed within 20 feet of the power cables being fed by these same breakers. The plant loads affected are the Unit 2 variable frequency drives and the Unit 3 recirculation motor-generator sets which provide power to drive the reactor recirculation pumps. As a result of this cable routing, fires in certain zones of the BFN plant could result in electrical faults on power cables which could not be de-energized by automatic breaker operation. Such faults could result in cable insulation fires being initiated in fire areas other than the area where the original fire occurred, thus creating an associated circuit of concern.
The apparent cause was an historical design error. Compensatory measures were implemented in accordance with the fire protection plan. The plant will be modified to correct the condition.
(If more space is required, use additional copies of (If more space is required, use additional copies of (If more space is required, use additional copies of (If more space is required, use additional copies of (If more space is required, use additional copies of NRC Form 366A) (17)
VII. ADDITIONAL INFORMATION
A.
Failed Components None B.
Previous LERs on Similar Events 260/94-002 260/96-01 R1 C.
Additional Information
None D.
Safety System Functional Failure Consideration:
The event under consideration did not involve an actual failure of any plant equipment, but rather it was a condition which, under certain postulated conditions, could have resulted in the failure of plant equipment. While certain postulated fire scenarios could have resulted in equipment losses beyond those previously analyzed, the loss of a safety system function was not specifically threatened.
This condition does not constitute a safety system functional failure as referenced in 10 CFR 50.73(a)(2)(v), and it will not be included in Performance Indicator reporting performed in accordance with NEI 99-02.
E.
Loss of Normal Heat Removal Consideration:
N/A This event did not involve a reactor scram.
VIII. COMMITMENTS
None