ML20135D709

Forwards Responses to Sections Iii.C & Iii.E of Generic Ltr 83-28,per NRC 850521 Ser.Responsibilities of Shift Supervisor & Sys Performance Group Discussed

ML20135D709
Person / Time
Site:	Farley
Issue date:	09/06/1985
From:	Mcdonald R ALABAMA POWER CO.
To:	Varga S Office of Nuclear Reactor Regulation
References
GL-83-28, NUDOCS 8509160160
Download: ML20135D709 (4)

Text

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" lling Addrxs Alabama Power Company 600 North 18th Street Post Office Bon 2641

. Dirmingham. Alabama 35291 Telephone 205 783-6090 R. P. Mcdonald Senior Vice President Fhntndge Building /\l[l!){llln)POWCf September 6,1985 Docket Nos. 50-348 50-364 Director, Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Mr. S. A. Varga Joseph M. Farley Nuclear Plant - Units 1 and 2 Additional Response to Generic Letter 83-28, Item 1.1 Gentlemen:

By letter dated May 21, 1985, the NRC issued to Alabama Power Company (APCo) a Safety Evaluation Report (SER) on Generic Letter 83-28, Item 1.1.

The SER concluded that the actions taken by APCo on Item 1.1 were acceptable with two exceptions. These exceptions were denoted as Sections III.C and III.E in the Evaluation and Conclusion Section of the SER. Attached is the APCo response to the two exceptions denoted in the referenced SER.

If there are any questions, please advis .

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/[~k LO R. P. Mcdonald RPM /RGW:ddb-D1 Attachment cc: Mr. L. B. Long Dr. J. N. Grace Mr. E. A. Reeves Mr. W. H. Bradford 8509160160 850906 PDR E ADOCK 05000348 PDR h0 l i

. Attachment Additional Response to Generic Letter 83-28,

, Item 1.1

. NRC Concern;Section III.C The licensee has not addressed the methods and criteria for comparing the avent information with known or expected plant behavior. We recommend that the pertinent data obtained during the post-trip review be compared to the applicable data provided in the FSAR to verify proper operation of the systems or equipment. Where possible, comparisons with previous similar events should be made.

APCo Response;Section III.C As discussed in the November 4,1983 APCo response to Section 1.1, the Shift Supervisor is responsible for reviewing the events relative to the trip to determine the cause and verify proper operation and response of the reactor and associated equipment. The results of this

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review and analysis are then discussed with the on-call Emergency Di rector. The Emergency Dire. or is responsible for determining if further review of the trip data is required and any immediate corrective actions which should be taken before the reactor can be restarted.

Attachment Page 2 The review performed by the Shift Supervisor and Emergency Director is

based on experience and training as well as the Emergency Response Procedures. The procedures provide

l 1) The symptoms, inmediate operator actions, and subsequent operator actions for a reactor trip.

2) Guidelines for operator action in the event a required reactor or turbine trip does not occur automatically.

3) Guidelines, along with appropriate portions of Critical Safety Function status trees, to ensure that the plant stays within design parameters and that the plant response to transients is as expected.

4) Guidance for restoring critical plant parameters to within design i

limits should these parameters exceed the limits.

5) Mitigating actions should plant parameters stay outside design j limits.

After the above reviews are performed, a Reactor Trip Form documenting the Shift Supervisor's review and analysis of the event is completed.

This form, along with the computer post-mortem printout, the alarm I

typewriter printout, and the sequence of events and incident reports, are reviewed by supervisory personnel in the Operations Group. The information contained in this package is not compared to the FSAR since the guidelines contained in the FSAR are based on worst case conditions. The Emergency Response Procedures provide guidance for expected plant behavior. Analyses of reactor trip transients are not

Attachment Page 3 included in the FSAR. The FSAR includes analyses for worse case transients. Also, comparisons with similar events are not made since no two trips are identical in nature and do not require the same plant responses.

NRC Concern;Section III.E The licensee has provided for our review a systematic safety assessment program to assess unscheduled reactor trips. We recomend that this program be revised to include the above cited methods and criteria for comparing the event information with known or expected plant behavior.

APCo Response;Section III.E The Systems Performance Group is responsible for performing an independent assessment of any event for the purpose of identifying the cause of the trip and verifying that proper corrective action (s) has been implemented or planned. This review verifies that the operation of critical components and the response of critical parameters were within guidelines as specified by the Emergency Response Procedures.

As indicated in the APCo response to Section III.C, a review of pre' ;ous events is not performed by the Systems Performance Group r:nce, in most cases, no two events are identical in nature and the plant response to any two events will be different.