ML17258A344

Forwards Comments on NRC Review of Operating Experience Non-DBE Reductions in Coolant Inventory (Leaks) Suggested as More Appropriate Categorization of Steam Generator Tube Leaks

ML17258A344
Person / Time
Site:	Ginna
Issue date:	11/25/1981
From:	Arthur J ROCHESTER GAS & ELECTRIC CORP.
To:	Crutchfield D Office of Nuclear Reactor Regulation
References
NUDOCS 8112020529
Download: ML17258A344 (22)

Text

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SJWgk'I ESN ROCHESTER GA'S AND ELECTRIC CORPORATION

~ 89 EAST AVENUE, ROCHESTER, N.Y. 14649 Sf *TE JOHN ARTHUR Vice Preeident end Chief Engineer TELEPHONE

• eEA cooE vie 546.2700 November 25, 1981 Director of Nuclear Reactor Regulation Attention:

Mr. Dennis M. Crutchfield, Chief Operating Reactors Branch No.

5 U.S. Nuclear Regulatory Commission Washington, D.C.

20555

Subject:

Review of Operating Experience R.

E. Ginna Nuclear Power Plant Docket No. 50-244

Dear Mr. Crutchfield:

We have reviewed the NRC-Contractor prepared assessment

,of the Ginna operating experience which was provided by your letter dated October 8, 1981.

Detailed comments are provided in Attachment 1 to this letter.

A number of our comments relate to categorization of steam generator tube leaks as tube failures.

Since leakage has been below O.l gallons per minute, we believe your, categorization under the category "Decrease in Inventory" to be inappropriate.

A more appropriate category is "Non-DBE Reductions in Coolant Inventory (Leaks)".

We suggest that, the report be revised accordingly.

,As identified in your report, the "C" safety injection pump emergency bus breaker has been subject. of a number of failures.

Since the period covered by your report.,

we believe we have identified and remedied the source of those malfunctions.

Attachment 2 to this letter describes the source of malfunction and the corrective actions taken.

Very truly yours, E. Arthur Attachments II 8ii2020529 8iii25 PDR ADOCK 05000244 P

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Attachment 1

Comments on NRC Report:

Ginna Operating Experience Pacae Comment 4-1 If steam generator tube leaks are treated as leaks, and not as failures, the number of DBEs entered is 18, not 23.

4-8 The number of DBEs should be 18, not 23.

4-9 The line entitled steam generator tube failures should be deleted and totals should be corrected.

4-11 Fuel/cladding concerns resulted in power operation being limited to 1266 MWt during much of 1972 and the first half of 1973

'-14 The repair of July 5, 1970 was to the pressurizer spray valve, not the pressurizer control valve.

4-16 At the top of the page, the block valve is upstream of the power-operated relief valve, not downstream.

4-16 A second refueling outage occurred in the fall of 1972 in which 48 fresh fuel assemblies were loaded.

This resulted in removal of all failed fuel from the reactor.

4-19 With respect to the low pressure

turbine, it should be noted that both low pressure turbine rotors have been replaced with rotors of an improved design.

In addition, a spare rotor is being stored on site.

4-21 The number of DBEs should be 18, not 23.

The phrase "five involved steam generator tube leaks" should be deleted.

4-22 Reporting requirements were revised on May 15, 1973 (in Change Number 8 to the Technical Specifications) and again on November 3,,

1975 (in Amendment Number 8 to the Operating

License, Change Number 17 to the Technical Specifications). 'aking these changes into

account, the statement that there is an upward trend in reportable events is incorrect and should be deleted.

Pacae 4-27 Comment The report does not explain the rationale behind the categorization provided in Table 4.6 for the loss of offsite power of 1973, particularly with respect to significance category S7.

4-28 Regarding the spurious closure of the MSIVs, it is,not clear that both MSIVs cl'osed simul-taneously due to flow impingement.

It is possible that one MSIV closed spuriously

and, due to the resulting pressure

wave, the other closed.

In addition, neither valve "failed".

In fact, both valves performed their desired safety function of closing.

Failure of the valve to close would, indeed, be a failure.

A report of the June 6,

1975 event was provided to the NRC in the Annual Report for 1975, submitted by letter dated February 27, 1976.

4-29 Malfunction of valve position indication in itself did not reduce the capability of the safety injection system since a safety injection signal would cause the valves to go to their desired position regardless of position indication.

4-34 We suggest that the length of an outage is not an appropriate criteria for categorization as "conditionally significant".

The length of an outage may be related only to commercial concerns and not to safety concerns.

Thus,'ategory C5 should be deleted.

4-34 The term "failed" steam generator tubes should be replaced with "leaking" steam generator tubes.

4-35, 4-36 Steam generator tube failures should be restated as steam generator tube leaks.

4-38 See attachment 2 to this letter for resolution of this issue.

4-43 Under report number 78-06, the last line of the description should read "snubbers" not "scrubbers".

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Pacae 4-44 Comment The fourth and fifth sentence of the second paragraph should be rewritten as follows to be consistent with current valve numbers and terminology:

"The hydrogen cooler tem-perature

rose, and th'e normal hydrogen cooler bypass valve (V-4229) closed.

This caused the condensate bypass valve (V-3959) to II open...

4-51, 4-54 The number of reportable events per year should be related to changes in reporting requirements (see th'e comment regaiding page 4-22).'-55 Control rod malfunctions have apparently been resolved (see the discussion contained on report page 4-37).

Safety injection pump C breaker problems apparently have been resolved although a subsequent failure has occurred (November 5, 1981).

Attachment 2 discusses the resolution of previous failures.

The most recent failure was due to a faulty lockout coil.

A periodic replacement program should resolve the lockout coil problem.

Item number 16 should refer to the pressurizer PORV block valve, not. the relief valve.

A-9 Item 5, paragraph A, second and third sentence should be revised to read:

"The normal hydrogen cooler bypass valve (V-4229) closed due to high hydrogen cooler temperature.

The condensate bypass valve (V-3959) opened..."

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Attachment 2.

Investigation of DB50 Breaker Failures Letter dated May 27, 1981 from Ronald C. Johnson, Westing-house to John H. Smith, RG&E.

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Interoffice correspondence dated May ll, 1981 from G.

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Daniels, RGEE, to Bruce

Snow, RGEE.

Interoffice correspondence dated June 18, 1981 from George S. Link, RG&E to Bruce A. Snow, RGSE.

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Westinghouse Electric Corporation 711 Exchange Street Box 887 Rochester New York 14603 r716) 232 4380 May 27, 1981

'r. John H. Smith Rochester Gas 5 Electric Corporation 89 East Avenue Rochester, New York 14649

Subject:

Your PO EG-15187 Our GO Ref. RH36655Y8, Evaluation of DB50 Breaker of S.O.

27Y2384 from the Ginna Nuclear Plant.

Dear John,

This letter is to summarize our findings at East Pittsburgh on April 22, 1981.

Me found that when the electrical lockout was energized there was no clearance between the electr ic lockout arm and the tr ip bar.

This caused the breaker to fail to close reliably.

There is a screw and locknut adjustment, that should be set for 1/32-1/16" clearance between the electric lockout arm and the trip bar when the lockout is energized.

Me corrected this adjustment and.the breaker subsequently. closed reliably.

This is a breaker manufacturing assembly adjustment that is not normally adjusted in the field.

Very truly yours, Ronald C. John on Sales Engineer RCJ:sm

ROCHESTER GAS AND ELECTRIC CORPORATION INTEROFFICE CORRESPONDENCE May ll, 1981

SUBJECT:

Results of Analyses and Testing Performed to Determine Cause of Failure of Certain W "DB-50" Circuit Breakers at Ginna Station (EWR 3073)

TO:

Bruce Snow Superintendent In order to confirm the findings of preliminary analyses and tests performed at the site, Breaker SIP1Cl was sent to the East Pittsburgh Plant of Westinghouse for diagnostic testing.

The result of these tests are described below.

'he failure of Circuit Breaker SIPlCl to close upon receipt of a close signal was determined to be due to inadequate clearance between the lockout assembly and the "tripper bar" which prevented the tripper. bar from locking into the closed position, and therefore simulated an overcurrent condition in the breaker.

It should be noted that the principal function of the tripper bar is to open the breaker when the overcurrent coil internal to the DB-50 breaker actuates (See figure 1).

Existing instruction books for the "DB" breakers with the lockout attachment state there are no adjstments for this clearance,

however, Westinghouse has since developed an adjustment method.

By loosening the

locknut, the elastic stop nut can be turned to adjust the clearance between the lockout assembly and the tripper bar.

The clearance should be 1/32"-1/16" with the lockout coil in the energized position (See figure 2).

Results of Anal ses and Testin Performed to Determine Cause of Failure of Certain W "DB-50" Circuit Breakers at Ginna.Station (EWR 3073)

It is recommended that plant maintenance procedures be modified to check, and adjust if needed, clearance of the lockout assembly on "DB" breakers that have this lockout feature in accordance with the W procedure.

This is in addition to any existing main-tenance procedures that may apply.

When maintenance procedures have been modified, reliability and availability of the breakers will be acceptable.

Installation of an auxiliary relay in lieu of the lockout assembly was considered, but rejected as the breaker interlock (designed to prevent paralleling busses 14 and 16) can then be defeated by manual closure.

For further assistance and information in this matter call John H.

Smith of the Electrical Engineering Group.

G.W. Daniels Manager, Electrical Engineering Attachment xc:

R. Smith J.

Smith G. Link J. StMartin R. Latz C. Edgar File/ENR 3073 Elec.

Eng. File 13Nl-RR-L0362

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- Electric Lockout Attactunent - Construction Details

ROCIiESTER GAS AND ELECTRIC CORPORATION INTEROFFICE CORRESPONDENCE June 18, 1981

SUBJECT:

Close out of ElfR 3073 DB-50 6

DB-75 Circuit Breaker Failures TO=

Bruce A. Snow Superintendent RE.

1) tIay 11, 1901 memo G.I/. Daniels to B.A. Snow

2) March 19, 1981 memo J. St.bIartin to R.E.

Smith The results of the testing performed by Westinghouse on the DB-50 breakers were described in reference 1.

The specific cause of the SIPlC1 breaker failures was determined to be the mechanical alignment of the lockout coil assembly and the tripper bar.

This problem was localized within the breaker mechanism itself and not related to external control wiring or equipment.

Corrective action was proposed and adjustments were made on all DB-50 breakers containing lockout coils.

A review of DB-75 circuit breaker failures, summarized in attachment 1,

has also been made.

These events have resulted principally from random component failures.

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EWR 3073 DB-50 R DB-75 Circuit Breaker Failure Since the DB-75 failure rate is not at this time unusually high, and does not exhibit any systematic failure mode, it is recommended that E$ 1R 3073 be closed out.

Surveillance of DB-75 breakers should, however, be continued and any new failures be brought to the attention of the Electrical Engineering Grouo.

Senior Electrical Engineer GSL:rh xc: G.iI. Daniels R.E.

Smith J. H. Smith L.S.

Lang R. Latz G. Larizza File/EWR 3073 Elec.

Eng. File 1331-RR-L0385

+V ql DATE OF EVENT/NRC REPORT 8

BREAKER PROBABLE FAILURE MODE 08/22/75 Unusual Event 75-5 lA Diesel Supply to Bus 16 Not apparent from review of event.

09/14/77 LER 77-19 1B Diesel Supply to Bus 16 Secondary contact finger bent.

08/16/78 LER 80-07 1B Diesel Supply to Bus 16 Bad connection at control power fuse block.

09/13/79 LER 79-18 1B Diesel Supply to Bus 16 Overcurrent relay lacked contin-uity.

09/10/80 LER 80-08 1B Diesel Supply to Bus 16 Binding of control relay anti-pump release lever guide pin.

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