ML20206J470
| ML20206J470 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 04/07/1987 |
| From: | Gridley R TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| EA-86-056, EA-86-56, NUDOCS 8704160102 | |
| Download: ML20206J470 (5) | |
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TENNESSEE VALLEY AUTHORITY CHATTANOOGA. TENNESSEE 374o1 SN 157B Lookout Place APR 071987 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C.
20555 Gent.lemen:
In the Matter of
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Docket Nos. 50-259 Tennessee Valley Authority
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50-260 50-296 BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 - SUPPLEMENTAL RESPONSE TO ENFORCEMENT ACTION EA-86-56 Enclosed is our response to G. G. Zech's March 4,1987 letter to S. A. White which concluded that our denial of Violation II from the subject Enforcement Action EA-86-56 was not supportable. Our original response on this item was transmitted to NRC in the October 8, 1986 letter from S. A. White to James M. Taylor. Violation II was a Level IV violation related to secondary 4
containment damper timing criteria.
Af ter careful consideration, we have decided to withdraw our denial'of this violation as explained in the enclosure. We appreciate your previous reconsideration of our position, and the discussion regarding the application of 10 CFR 50.59 on page 2 of your letter.
Please refer any further questions to M. J. May, Browns Ferry Site Licensing, at (205) 729-3566.
To the best of my knowledge, I declare the statements contained herein are complete and true.
Very truly yours, TENNESSEE VALLEY AUTHORITY b
R. Crldley, Director Nuclear Sa ty and Licensing Enclosure ec:
See page 2
'1 G704160102 870407 2
PDR ADOCK 05000259 I
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'y 1 g An Equal Opportunity Employer
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. U.S. Nuclear Regulatory Connission APR 071987 cc (Enclosure):
Mr. G. G. Zech, Assistant Director Regional Inspections Division of TVA Projects office of Special Projects U.S. Nuclear Regulatory Consission Region II 101 Marietta St., NW, Suite 2900 Atlanta, Georgia 30323 Browns Ferry Resident Inspector Browns Ferry Nuclear Plant P.O. Box 311 Athens, Alabama 35611
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SUPPLEMENTAL RESPONSE ENFORCEMENT ACTION REPORT BROWNS FERRY NUCLEAR PLANT (BFN) 50-259/86-56, 50-260/86-56, AND 50-296/86-56 DR. J. NELSON GRACE'S LETTER TO S. A. WHITE DATED SEPTEMBER 8, 1986 Violation Not Assessed a Civil Penalty Item II 10 CFR 50.59(a) allows the holder of a license to make changes in the facility as described in the Final Safety Analysis Report (FSAR) without prior Commission approval unless it involves a change to the Technical Specifications or is an unreviewed safety question.
An unreviewed safety question is created if the consequences of an accident or the malfunction of the equipment important to safety previously evaluated in the FSAR may be increased.
10 CFR 50.59(b) requires that the licensee maintain records of changes in the facility to the extent that such changes constitute changes in the facility as described in the FSAR. These records shall include a written safety evaluation which provides the bases for the determination that the change does not involve an unreviewed safety question.
Contrary to the above, the licensee changed the facility as described in the FSAR without having determined whether the change involved an unreviewed 1
safety question.
On November 25, 1985, the licensee changed the acceptable closure time of secondary containment isolation dampers from two seconds as specified in Section 5.3.4.2 of the FSAR to 10 seconds.
Although the licensee performed an analysis, it did not determine whether the change involved an unreviewed safety question. The change was implemented by the licensee through an internal memorandum pending a change to the FSAR to be submitted as part of the routine annual update per 10 CFR 50.71.
This is a Severity Level IV violation (Supplement I).
1.
Admission or Denial of the Alleged Violation TVA admits this violation.
2.
Reasons for the Violation Our October 8,1986 response to EA-86-56 presented a lengthy discussion of the circumstancer related to this violation and will not be repeated here.
After careful consideration of the facts related to this vloistion, we have determined two causes. The first cause is specific to this coue and the second cause is related to more general principles regarding the appilcation of 10 CFR 50.59.
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. Cause 1 As explained in our October 8,1986 response to this item, both the FSAR description of the integrated operation of the Standby Cas Treatment System (SGTS)/Socondary Containment Isolation System and subsequent engineering evaluations strongly supported that the accident analysis (fuel handling accident) was only sensitive to the speed of the refueling zone exhaust dampers. We point out that the refuel zone exhaust dampers met the FSAR closing speed critoria. The plant staff concluded that the FSAR wording was incorrect with regard to damper nomenclature by virtue that the reactor and refuel zone dampers were imprecisely categorized together as " reactor zone" dampers with no clarification as to their functional performance requirements.
It was decided to clarify the FSAR during the annual updato to more cicarly distinguish the difference between the functional performance characteristics of the reactor and refuel zone dampers. This proposed change was considered a clarification change and, therefore, was categorized as a routino item to correct during the yearly update. The November 25, 1985 memorandum referred to in the violation description did not authorizo changes to be made in the closure speeds of the dampors, but rather confirmed that the as-found speeds were acceptable.
Cause 2 The second cause is related to broador considerations in rosard to the use of 10 CFR 50.59 in evaluating changes to the facility.
It was our previous understanding that 10 CFR 50.59 was intended for uso in evaluating licenseo (planned) changes to the facility and not for justifying nonconformances betwocn the FSAR and the physical facility.
NRC's reasoning in this matter is shown on page 2 of the onclosure to the March 4, 1987 NRC Ichter as follows:
" Alternately, the noncompilanco may be evaluated and found to be acceptable as is.
Should this second option be selected, no change would be made to the system or components' as found condition, howevor, a change to the system as described in the FSAR would have been doeund acceptablo and mado by default.
The licensee is required by 10 CFR 50.59 to dolormine whether an unroviewed safety question is involved prior to changing the facility as described in the FSAR".
We have enrofully examined this roanoning and we will modify our procedural methodology for the application of 10 CFR 50.59 to include this alternativo in addressing nonconformances. Our previous understanding with regard to the rigorous application of 10 CFR 50.59 contritated to tho subject failure to perform a 10 CFR 50.59 since there was not a (planned) physical change to tho facility.
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Corrective Steps Which Have Been Taken and Results Achieved As noted in our October 8, 1986 response, a formal 10 CFR 50.59 analysis was subsequently performed and the FSAR revised to include a new fuel handling accident analysis.
4.
Corrective Steps Which Will Be Taken to Avoid Further Violations The applicable Program Manual Procedure, PHP 0604.04, " Evaluation of~
Changes Tests, and' Experiments," and the Nuclear Quality Assurance-Manual (NQAM), Section 2.16 " Corrective Action," will be revised to require an evaluation in accordance with PHP 0604.04 before closure of a Condition Adverse to Quality Report (CAQR) that deals with a discrepancy between the plant or its procedures and the FSAR.
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Date When Full Compliance Will Be' Achieved The Program Manual Procedure and NQAM will be revised by June 30, 1987.
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