ML20237K704

Monthly Operating Rept for May 1987

ML20237K704
Person / Time
Site:	Oyster Creek
Issue date:	05/31/1987
From:	Baran R, Fielder P, Sedar J GENERAL PUBLIC UTILITIES CORP.
To:	NRC
References
NUDOCS 8708190376
Download: ML20237K704 (9)

Text

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4 MONTHLY OPERATING REPORT - MAY 1987 At the beginning of the report period, Oyster Creek was in an extended shutdown to replace the cooling coils on four (4) drywell recirculation fans.

The plant was initially shutdown on April 24 to replace an acoustic monitor associated with an electromatic relief valve.

Following completion of repairs, reactor startup commenced on May 14.

The generator was placed on-line on May 16 and reactor power was slowly increased.

On May 19, power was reduced to approximately 360 hMe to repair an oil leak on a reactor feed pump.

Power was subsequently increased to 558 MWe, Maximum generator load was achieved on May 20.

Full power was maintained for the balance of the report period except for one (1) brief power reduction to 575 AMe to perform turbine valve testing.

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MONTHLY OPERATING REPORT MAY 1987 The following Licensee Event Reports were submitted during the month of May 1987:

Licensee Event Report 50-219/87-019 - " Limiting Safety System Setpoint for Total Recirculation Flow Exceeds Tech Specs Due to Personnel Error" On April 3,1987 during a maintenance check and on April 6,1987 during a surveillance test, the reactor recirculation flow scram, setpoints for Reactor Protection Systems I and II (RPS I ant II), respectively, were found to be above the Limiting Safety System setpoint of 117%.

The apparent cause of both (RPS I and RPS II) scram setpoints being out of specification was personnel error in that they were apparently inadvertently set high during a surveillance test performed March 30, 1987.

A recirculation flow scram would not have occurred within requisite flow limits during the time the setpoints were out of specification.

However, the APRM power scram setpoint of approximately 114.5% would have scrammed the reactor prior to safety limits being exceeded in the event of a recirculation flow increase transient.

The technician responsible for this event will have his qualification reviewed and his previous work will be checked.

The surveillance procedure will be revised to specify the use of a more accurate flowmeter for setpoint verification.

Instrument technicians will be required to read a critique of this event and this LER.

Further, they will be required to read the revised procedure for the proper techniques for setpoint verification and adjustment.

Licensee Event Report 50-219/87-020 " Technical Speci fication Surveillance Overdue due to Inadequate Shif t Turnover Caused by Personnel Error" On April 2,1987 on the 1600-2400 hours shif t, the daily operability check of the main steam isolation valves (MSIV) was not performed within the time frame as specifled in the Technical Specifications.

The Group Shift Supervisor (GSS) intentionally postponed the 5% MSIV closure test but failed to document the postponement in the control room logs and turnover documents.

The postponement was warranted pending investigation of an unusual noise originating in the vicinity of the high pressure turbine.

At shift turnover the GSS did not i nform the oncoming shift of the postponed surveillance nor was this information documented in any of the logs or turnover checklists.

The GSS was counselled by Operations l

management, the incident will be discussed with each shift during the weekly plant status upda te meeting with Operations management, and discussion of the incident will be included in regular licensed operator training.

In addition, required reading will be issued to operators to emphasize the need and requirements for documenting the postponement of l

surveillance tests.

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l'icensee' Event Reports May 1987 Page 2 i

Licensee Event Report 50-219/87-021

" Technical Specification Violation.

Caused by Blocking Open Containment Vacuum Breakers Due to Personnel Error" On April 24, 1987 at approximately 0330 hours0.00382 days <br />0.0917 hours <br />5.456349e-4 weeks <br />1.25565e-4 months <br /> two torus to drywell vacuum breakers were blocked open to assist containment deinerting when primary i

containment integrity was required.

The plant was being shutdown when the vacuum breakers were blocked open with the mode switch in RUN, reactor power at 23% (450 MWt) and reactor temperature at 530*F.

The Operations Shift Supervisor originated the Temporary Variation to block open the h

vacuum breakers, and both he and the Shift Technical Advisor performed a safety review which did not reveal a safety concern.

Within approximately 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> the Operations Shift Supervisor realized his error and removed the Temporary Variation.

Had a Loss of Coolant Accident occurred when the vacuum breakers were blocked open, primary containment integrity would have been compromised.

The cause of this event is personnel error in that personnel made a cognitive error in evaluating the safety significance of L

this temporary variation.

A contributing factor was improper implementation of the safety review process in regard to temporary uriations.

Corrective action consisted of revising procedures to require a more detailed and documented safety review with written justification for safety significance determinations.

Key site personnel responsible for performing safety reviews were retrained on these procedure changes prior to restart.

Training for all site personnel involved in the process will be conducted.

All existing temporary variations were reviewed for safety review adequacy.

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Licensee Event Re3 ort 50-219/87-022 " Plant Shutdown Required by Inoperable Acoustic Monitor

)ue to Marginal Splice Design Resul ting in Cable Damage During Installation" On April 24, IJ7 a plant shutdown was completed as required by Technical Specifications because of an inoperable electromatic relief valve position indicator.

Before the shutdown, t ae plant was operating at approximately 99% power.

The position indication (acoustic monitor) became inoperable on April 22, 1987 at 1220 hours0.0141 days <br />0.339 hours <br />0.00202 weeks <br />4.6421e-4 months <br /> and subsequent troubleshooting efforts indicated an open circuit had occurre'd inside the drywell.

The cause of the event was the failure of a cable splice at the connection between a coaxial cable and one wire of a twisted shielded pair.

The splice had been installed on October 13, 1986 in accordance with the plant's Equipment Qualification program.

The safety significance is minimal because the backup position indicator remained operable.

Short term corrective action was taken to replace the failed splice and i ts I

associated acoustic monitor, and test all other acoustic monitors.

Two other channels (one electromatic relief valve and one safety valve) were repai red due to weak output identified during the testing.

A spare acoustic monitor for each channel has been installed in the drywell to be used in the event of another channel failure.

Long term corrective action is being evaluated to improve th? design of the spliced connection to minimize weak points in the splice assemblies, or eliminate the splice by using different qualified connectors, l

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(icenseeEventReports May 1987 Page 3 Licensee Event Report 50-219/87-023 " Partial Primary Containment Isolation During Testing Due to Procedural Inadequacy" On April 30, 1987 while performing surveillance testing of reactor protection motor generator sets, a partial primary containment isolation occurred.

The root cause of the event was procedural inadequacy.

The surveillance procedure had not been revised to reflect two modifications to the plant, each of which would have caused a partial isolation independent of the other.

In both cases the review for procedures that could be affected by the modification failed to identify this procedure as requiring changes.

In at least one occurrence in November 1986, this test had been performed with a partial isolation occurring.

It is possible that two other events occurred in October and November 1984.

None of these were noted in the procedures or operator logs.

The a f fected procedures and both modifications will be reviewed to ensure both modifications are reflected where necessary.

Notification requirements have been emphasized to shift supervisory personnel.

This report will be required reading for engineering and operations personnel.

Guidance will be provided to improve engineering' reviews of modifications.

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OPERATING DATA REIORT OPERATING STATUS 1.

DOCKET:

50-219 2.

REPORTING PERIOD:

MAY, 1987 3.

UTILITY COtiTACT:

JOHti H. SEDAR, JR.

609-971-4698

4., LICEllSED TFERMAL POWER (Mdt):

1930 5.

NAMEPLATE RATING (GROSS Mde):

687.5 X 0.8 = 550 6.

DESIGN ELECTRICAL RATING (tiET Mde):

650 7.

MAXIMUM DEPENDABLE CAPACITY (GROSS Mde):

650 8.

MAXIMUM DEPEtIDABLE CAPACITY (NET Kde):

620 9.

IF CHAllGES OCCUR ABOVE SINCE LAST REPORT, GIVE REASOt1S:

NOt1E 10.

POWER LEVEL 'IO WHICH RESTRICIED, IF ANY (NET Kde):

N/A 11.

REASON FOR RESTRICTION, IF ANY:

NOt!E MONTH YEAR CUMULATIVE 12.

REPORT PERIOD HRS 744.0 3623.0 152856.0 13.

HOURS RX CRITICAL 406.8 2364.8 97201.3 14.

RX RESERVE SHTDWN HRS 0.0 0.0 918.2 15.

HRS GENERATOR Oti-LINE 377.5 2241.3 94611.3 16.

UT RESERVE SHTDW!l HRS 0.0 0.0 1208.6

17. GROSS THERM ENER (MWH) 679700 3895304 156851689

18. GROSS ELEC ENER (MWH) 226740 1319940 52988185

19. NET ELEC ENER (MdH) 216156 1261592 50871669 20.

UP SERVICE FACIOR 50.7 61.9 61.9 21.

Vf AVAIL PACIOR 50.7 61.9 62.7 i

22.

UT CAP FACIOR (MDC NET) 46.9 56.2 53.7 23.

UT CAP FACIOR (DER NET) 44.7 53.6 51.2 24.

UT FORCED OUTAGE RATE 49.3 38.1 11.3 25.

FORCED OUTAGE HRS 366.5 1381.7 12033.5 26.

SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, DURATION):

N/A

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1 27.

IF CURRENTLY SHUTIW1 ESTIMATED STARTUP TIME:

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Docket No. 50-219 REFUELING INFORMATION - MAY, 1987 Name of Facility: Oyster Creek Station #1 Scheduled date for next refueling shutdown: N/A Scheduled date for restart following refueling:

Will refueling or resumption of operation thereaf ter require a Technical Specification change or other license amendment?

No I

Scheduled date(s) for submitting proposed licensing action and supporting information:

t Important licensing considerations associated with refueling, e.g.,

new or different fuel, design or supplier, unreviewed design or performance analysis methods, significant changes in fuel design, new operating procedures:

1. General Electric Fuel Assemblies - fuel design and performance analysis methods have been approved by the NRC.

2. Exxon Puel Assemblies - no major changes have been made nor are there any anticipated.

The number of fuel assemblies (a) in the core 560

=

(b) in the spent fuel storage pool = 1392 (c) in dry storage 20

=

l The present licensed spent fuel pool storage capacity and the size of any 2

increase in licensed storage capacity that has been requested or is planned, in number of fuel assemblies:

Present licensed capacity:

2600 The projected date of the last refueling that can be discharged to the spent 3

fuel pool assuming the present licensed capacity:

1 Retacking of the fuel pool is in progress. Six (6) out of ten (10) racks have been installed to date.

When reracking is completed, discharge capacity to the spent fuel pool will be available until 1990 refueling outage.

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r AVERAGE DAILY POWER LEVEL t1ET MWe DOCKET f........

50-219 UNIT........... OYSTER CREEK #1 REPORT DATE.

. JUNE 2, 1987 COMPILED BY.

.... JOHN H. SEDAR, JR.

TELEPHONE #......

609-971-4698 k

MOtiTh MAY, 1987 DAY FM DAY Ffd 1.

0 16.

155 2.

0 17.

468 3.

0 18.

494 4.

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499 5.

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630 8.

0 23.

630 9.

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630 11.

0 26.

623 12.

0 27.

633 13.

0 28.

631 14.

0 29.

627 15.

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618 31.

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1 Post Office Box 388 Route 9 South Forked River, New Jersey 087310388 609 971-4000 Writers Direct Dial Number:

Director June 15, 1987 Office of Management Information U.S. Nuclear Regulatory Commission Wa shing ton, 'DC 20555

Subject:

Oyster Creek Nuclear Generating Station Docket No. 50-219 Monthly Operating Report i

In accordance with the Oyster Creek Nuclear Generating Station Operating 1

License No. DPR-16, Appendix A, Section 6.9.1.C, enclosed are two (2) copies of the Monthly Operating Data (gray book information) for the Oyster Creek Nuclear Generating Station.

If you should have any questions, please contact Mr. Joseph D. Kowalski, Oyster Creek Licensing Manager at (609)971-4643.

Very truly yours, u/

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Vice President and Director Oyster Creek PDF:KB:dmd(0841 A)

Enclosures cc:

Director (10)

Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission l

Washington, DC 20555 Mr. William T. Russell, Administrator Region I l

U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406 Mr. Alexander W. Dromerick, Project Manager j

U.S. Nuclear Regulatory Commission Division of Reactor Projects I/II 7920 Norfolk Avenue, Phillips Bldg.

Bethesda, MD 20014 I

i NRC Resident Inspector

/

Oyster Creek Nuclear Generating Station I

g GPU Nuclear Corporahon is a subsidiary of the Genera: Pubhc Utihties Corporation

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