ML17249A259
| ML17249A259 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 11/07/1979 |
| From: | White L ROCHESTER GAS & ELECTRIC CORP. |
| To: | Ziemann D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17249A260 | List: |
| References | |
| NUDOCS 7911140298 | |
| Download: ML17249A259 (8) | |
Text
SUBJECT:
Responds to 790921 ltr re multiple equipment failures L
surveillance testing errors. Surveillance L calikr procedures r'equire notification to on duty shift foreman L heed control operators DISTRIBUTION CODE:
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TITLE: General D>stribution for after Issuance of Opei ating Lic NoTEs:3+.~ ~d,o"tL6ara Z ~~~~~<~E.
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05000244 AOTH<NAME AUTHOR AFFILIATION HHITEiL~Do Rochester Gas 8 Electric Corp.
REC IP ~ NAME RECIPIENT AFFILIATION ZIEHANNgD ~ LE Opeiating Reactors Branch 2
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'OCHESTERGAS AND ELECTRIC LJalONfk&O+I CORPORATION o
89 EAST AVENUE, ROCHESTER, N.Y. I4649 LEON D. WHITE, JR.
VICE PRCSIOENT TELEPHONE AREA cooc TI6 546-2700 November 7, 1979 Director of Nuclear Reactor Regulation ATTN: Mr. Dennis L. Ziemann, Chief Operating Reactors Branch No.
2 U. S. Nuclear Regulatory Commission Washington, D. C.
20555
Subject:
Multiple Equipment Failures and Surveillance Testing Errors R. E. Ginna Nuclear Power Plant, Unit No.
1 Docket No. 50-244
Dear Mr. Ziemann:
In response to Mr. Eisenhut's letter of September 21, 1979, received on October 8, 1979, concerning multiple equipment failures and surveillance testing errors, the following information is provided.
An investigation has been made of all reactor trips at Ginna Station.
The only pertinent trip occurred on 7anuary 3, 1970 with a reactor trip and safety injection.
The trip was initiated when steam pressure channel 478 was isolated due to a sensing line leak while steam pressure channel 483 was undergoing calibration by the I&C department.
It should be noted, how-ever, that all systems performed as designed and no multiple equipment failures resulted.
As a result of this incident the following administrative and procedural changes were implemented.
Surveillance and calibration procedures now require notification not only to the on-duty Shift Foreman but also to the Head Control Operator at the start and upon completion of the tests.
Certain surveillance procedures requiqe the test personnel to be relieved of all simultaneous responsibilities reIative to normal duties and routine operations while performing th'e test.
Their responsibilities will be covered by other onsite personnel.
Other sur-veillance procedures prohibit the performance of other activities related to maintenance or testing which may distract control room personnel involved in the performance of the test.
The calibration and surveillance activities on safety related systems are performed on a periodic basis.
This continual testing of safety related
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DATE November 7, 1979 To Director of Nuclear Reactor Regulation ATTN: Mr. Dennis L. Ziemann, Chief SHEET NO.
2 systems protects against the possibility of multiple equipment failures.
Implementations of these modifications to the administrative practices, surveillance and calibration procedures willavoid possible challenges to the features protecting against the possibility of multiple equipment fail-ures at Ginna Station.
Supervisors of the operations, instrumentation and control, results and test, and training sections have been requested to review this occur-ence and stress the importance of avoiding challenges to the reactor pro-tective features.
Very truly yours,
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TRIBUTION KMQNW6ÃKG'itMPE Hr. Leon 0. V,hite, Jr.
Vice President Electric and Steam Proluction Rochester Gas and Electric Corporation 89 East Avenue Rochester, New York 14649 NRC'PDR Local PDR ORB ¹2 Reading NRR Reading DLZiemann HSmith JJShea PWagner OELD OISE (3)
DBrinkman JRBuchanan TERA ACRS (16)
NOV og ]979
Dear Hr. Rhite:
The staff has recently completed a review of the LER's and Technical Specification r'equirements related to the Control Rod Position Indication Systems (RPI) at Westinghouse PMRs.
He have determined that a wide varia-tion exists in the number of LERS received and the technical specification requirements and have, therefore, decided to clarify our'equirements.
At the time of development of the Standard Technical Specifications, a
systematic attempt was made to clarify potentially ambiguous specifications.
One such specification was the control rod misalignment specification for ltestinghouse-designed reactors.
1<estinghouse has performed safety analyses for control rod misalignment up to 15 inches or.24 steps (one step equals 5/8 inch).
Since analysis of misalignments in excess of this amount have not been submitted, we have imposed an LCO restricting continued operation with a misalignment in excess of 15 inches.
Because the analog control ro'd position indication system has an uncertainty of 7.5 inches (12 steps),.when an indicated deviation of 12 steps exists, the actual misa'lignment may be 15 inches.
This is because one of the coils, spaced at 3.75 inches, may be failed without the operator knowing about it.
The Standard Technical Specifications were written to eliminate any confusion about this, and allow a deviation of up to 12 indicated steps.
Surveillance requirements, on the indication accuracy of 12 steps were also prepared to ensure that the 15 inch LCO is met.
a There is no difference intended in requirements issued for any Hestinghouse reactor.
Mestinghouse has informed the NRC that all of their customers have been informed of this and that all the licensees should be following the same procedures regardless of the language of their Technical Specification.
That is, plants with Technical Specifications written in terms of 15 inch misalign-ment should be considering the 12 step instrument inaccuracy when monitoring rod position.
A related problem is that the installed analog control rod position indicating system equipment may not, in some areas, be adequate to maintain the control rod misalignment specification requirement because of drift problems in the D)PFN$$
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