ML20237K889
| ML20237K889 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 08/28/1987 |
| From: | Tucker H DUKE POWER CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8709080090 | |
| Download: ML20237K889 (5) | |
Text
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DUKE POWER GOMPANY P.O. nox 331(59 CHARLOTTE, N.C. 28242 HAL H. TUCKER reternoxe vmaPunainswt (704) 373-41538 NtM> LEAS PRODt'UTION
' August 28, 1987 U. S. Nuclear. Regulatory Conunission Attention: Document Control Desk Washington,'D C. 20555 Subj ect: Catawba Nuclear Station Docket Nos. 50-413 and 50-414 RII:PKV/MSL Violation 413/87-20-03
Dear Sir:
Attached is our response and denial of the subject Violation which concerns the thoroughness of an incident investigation at Catawba.
Very truly yours, I
s <._
Hal B. Tucker LTP/87/sbn Attachment xc:
Dr. J. Nelson Grace Regional Administrator U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Mr. P. K. Van Doorn NRC Resident Inspector Catawba Nuclear Station 8709080090 870828 0[
PDR ADOCK 05000413 G
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Violation:
Techt.ical Specification 6.8.1 requires that written procedures
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shall he established, implemented, and maintained covering the applicable procedures recommended in Appendix A to Regulatory l
Guide 1.33, Revision 2, February 1978.
Duke Pouer Company Safety Review Group Incident Investigation and Report ? reparation, SRG/2, and Catawba Nuclear Station Directive 2.8.1.
Problem Investigation Process and Regulatory Reporting collectively require reportable events to be thoroughly investigated, the cause determined, and the full implications of recurring events evaluated.
Contrary to the above, the licensee completed Incident Investigation Report C87-040-1 of May 14, 1987, and did not thoroughly investigate, determine the cause, and evaluate the full recurring event implications for valve INSPT5040 inappropriately remaining shut from April 7 to April 24, 1987, causing Containment Pressure Channel IV to be inoperable.
Response
Duke Power Company denies the alleged violation.
Duke Power Company has implemented written procedures to ensure that reportable events are thoroughly and consistently investigated.
Safety Review Group Procedure 2, Incident Investigation and Report Preparation, and Catawba Nuclear Station Directive 2.8.1, Problem Investigation Process and Regulatory Reporting, collectively address the investigation process, provide cause codes to be utilized, and require that the full implications be considered for recurring events.
The investigation performed to produce Incident Investigation Report C87-040-1, Rev.
O, and the subsequent LER 413/87-18 was performed in accordance with 10CFR 50.73, NUREG 1022, Station Directive 2.8.1, and Safety Review Group Procedure 2.
Reg Guide 1.33, Revision 2, February 1978, Appendix A, requires that written procedures shall be established to cover safety related activities during operations.
Appendix A does not address investigative activities performed by an Independent Safety Engineering Group or activities performed to satisfy the requirements of 10CFR 50.73, LER content.
Response
, Page 2 In that the incident was instrument related, the Safety Review Group investigator assigned was the group member most experienced in that area.
Concurrent with the Safety Review Group review, informal reviews were performed by a station Instrument technician and an Operations Staff Engineer.
The Operations Engineer could identify no operating or test _ag activity which would have manipulated the instrument root valve.
The technician could identify no maintenance activity which would have manipulated the root valve in the proximate time frame in which the closed valve was discovered.
In addition, station management was concerned that the actual cause had not been identified at the completion of the investigation.
Accordingly, the Safety Review Group Supervisor reviewed the Testing Logbook and the Tech l
Spec Equipment Out of Service Logbook for both Units in the suspected time frames in an attempt to identify the cause of the event.
In review of the efforts exerted, the depth of the investigation was considered reasonable and consistent with the guidelines of NUREG 1022 and Supplements.
As stated in the associated reports, the most appropriate source of information was reviewed.
During the review of the strip chart of the specific containment pressure transmitter output, it was not recognized by the investigator nor the technician that the transmitter was not tracking following calibration and independent verification of return to service on April 7, 1987.
Slight variations in the strip chart were noted following April 7, and were incorrectly concluded as indication that the transmitter was inservice.
Those variations were later determined to be the result of ambient temperature changes in the instrument's environment.
The investigator and technician failing to recognize the transmitter not being returned to service following calibration was the result of engineering judgement.
They did not recognize that more variance in the transmitter output would be expected during normal operation.
Supervision ensured that the most appropriate information sources had been reviewed and that the investigator was satisfied with his findings.
The previous similar incident was again reviewed during this investigation but could be concluded similar only in effect, not cause, due to dissimilarity of the sequences of events prior to the channels being found out of service.
As a followup to the Resident Inspectors' concerns that this was an apparent repeated event, a new investigator in the safety Review Group was assigned to further review both incidents.
Consequently, the inspection of the valve revealed a loose and mispositioned handwheel on the valve stem, allowing the handwheel to rotate without stem movement.
This apparently misled the technicians performing the calibration on April 7,
into believing they had opened the valve when in fact the valve remained closed.
Response
page 3 Following discovery of the slipping handwheel, all Unit 1 and 2 containment pressure channels were inspected and valve handwheels were verified to be properly installed and tightened.
To ensure that containment pressure channels are properly returned to service following calibration, Units 1 and 2 Calibration procedures were revised to specify improved methods of return to service verification.
Also, technician training will be provided to appropriate personnel concerning the proper operation and/or maintenance of Dragon instrument isolation valves.
j 1
The program has been reviewed in light of these two incidents and found to be satisfactory.
No changes could be identified which would improve effectiveness of the program.
The use of the Unknown cause in Catawba LERs is not considered inordinate.
In addition, this is the first known instance in which there was a misidentification of an event cause.
Catawba LERs have consistently been revised to provide new information when such information significantly affects the content of a previously submitted LER.
The need to thoroughly review indicants which appear to be repetitive has been reemphasized to the Safety Review Groups at Catawba, McGuire, and Oconee Nuclear Stations.
In support of this conclusion, the AEOD Assessments of Catawba LERS rated the overall quality as 7.9 compared to the industry average of 6.6 in the March 1984 to September 1985 time period, and a rating of 8.3 compared to the industry average of 8.0 in the October 1985 to September 1986 time period.
In addition, results of the July 1987 NRC Audit of Quality Verification Programs at Catawba reviewed the incident investigation process and selected reports and identified no concerns in that area.
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