ML20245B308

Monthly Operating Rept for Mar 1989 for Oyster Creek Nuclear Generating Station

ML20245B308
Person / Time
Site:	Oyster Creek
Issue date:	03/31/1989
From:	Fitzpatrick E, Sedar J GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 8904260011
Download: ML20245B308 (9)

Text

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MONELY OPERATING REIORT - MARCH 1989 e

l At the beginning of the report period, Oyster Creek was shut down for the Cycle 12 Refueling Outage.

On March 27, 1989, while in the start-up sequence, ic was observed that a recirculation pump mechanical seal was leaking.

Several hours later, after increasing pump speed, increased leakage was observed and the pump was exhibiting excessive vibration.

The plant. was shut down to accommodate pump disassembly / inspection and remained out-of-service at the end of the report period.

i 8904260011 890331

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MONTHLY OPERATING REPORT MARCH 1989 The following Licensee Event Reports'were submitted during tFe month of March 1989:

LER 89-003:' REACTOR ISOLATION DURING SURVEILLANCE TEST CAUSED BY FAILED FUSE On February 13, 1989 a reactor isolation signal was inadvertently initiated during a routine surveillance test. This.is considered reportable in accordance with 10CFR50.73(a)(2)(iv). The cause 'of the reactor' isolation signal 'is understood. A failed fuse in Reactor Protection System (RPS)

- Channel 2 caused several containment and reactor isolation relays to be in a

. hen another half trip reactor isolation signal W

deenergized or tripped state.

was introduced during the surveillance test in RPS Channel 1, a full reactor

-isolation signal was generated. The time and cause of the fuse failure could not be determined. The root cause of this event has been attributed to a human engineering design deficiency. The conditions that cause a half trip, as well as the pitysical results of a full isolation signal are annunciated.

There are, however, no alarms associated with a half or full isolation signal from the isolation relays themselves. A modification to provide a better indication of the status of half and full isolation signals'will be evaluated.

The safety significance of this event was minimal since this event occurred while the plant was shut down with the result being an unnecessary challenge to the reactor protection system.

. LER 88-004: MAIN STEAM ISOLATION SIGNAL DURING REACTOR PROTECTION SYSTEM POWER SUPPLY TRANSFER CAUSED BY INADEQUATE PROCEDURE On February 22,1989, at 0240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br />, a main steam isolation valve automatic closure signal occurred during transfer of a reactor protection system (RPS) power supply. At the time, the plant was in cold shutdown with reactor coolant temperature at 182*F. The cause of the event is procedural inadequacy, precipitated by inadequate communication in that a jumper intended to bypass a main steam line low pressure isolation signal did not accomplish its intended function.

In addition, the procedure for the RPS power supply l

transfer did not acknowledge that an isolation could be caused by the transfer evolution. This event had no safety significance.

Immediate corrective actions were taken to open isolated valves and revise the procedure for jumper placement so that the jumper will prevent an isolation signal upon loss of RPS power. Future corrective actions include revising the RPS power supply operating procedures and providing this information to procedure preparers and j

reviewers.

MONTHLY OPERATING REPORT MARCH 1989 (cont'd)

LER 89-005:

STANDBY GAS INITIATION DUE TO INADEQUATE SYSTEM DESIGN On February 17, 1989 at approximately 0242 hours0.0028 days <br />0.0672 hours <br />4.001323e-4 weeks <br />9.2081e-5 months <br />, while the reactor was in the refuel mode, the Standby Gas Treatment Sys ;em (SBGT.S) and a partial primary containment isolation were initiated when a electrical short circuit occurred inside a control room panel. At the time of the occurrence, the Operations Department was performing control rod insertion time tests and the short occurred when a jumper being used for the test slipped off its connection point. When this happened, a portion of the Reactor Protection System (RPS) was shorted, causing a protective fuse to blow, deenergizing a portion of the RPS circuit and initiating the event. The fuse was replaced, the isolation signal was reset, and the SBGTS and isolation valves were returned to normal.

The cause of the event has been attributed to the fact that the system is not designed to facilitate testing and jumpers must be used to provide connections between test points.

Because the test points were not designed to accommodate these jumpers, these jumpers could easily become dislodged from the screws to which they were attached. The safety significance of this event is considered minimal since the RPS circuit affected failed in the safe condition and all protection circuits operated as designed. Jumper construction will be

. evaluated and the best jumper for this application will be designated for use during this surveillance. The adequacy of the test point connections for this surveillance will also be evaluateo as part of a test point review already being conducted.

LER 89-006: TECH SPEC INCONSISTENCY RESULTS IN LESS THAN REQUIRED NUMBER OF AUTOMATIC DEPRESSURIZATION CHANNELS On February 24, 1989, a Technical Specification inconsistency was identified which had resulted in violations of the minimum number of operable channels required for the Automatic Depressurization System (ADS) while maintaining compliance with the limiting condition for operation of the Core Spray Systems. The safety significance of these events is considered minimal as: 1) at least one trip system of ADS capable of fulfilling the ADS function was available at all times; 2) logic inputs to the ADS are alarmed in the control room ensuring operator awareness to changing plant conditions; and 3) plant procedures provide guidance to the operator to inhibit the automatic function of the ADS during emergency conditions and depressurize the reactor in a more controlled manner.

The lack of correlation in operability requirements has been included in the Oyster Creek Technical Specification improvement program, and the procedure on evaluating Preliminary Safety Concerns has been revised to ensure more effective reviews. Plant management will review these operability requirements with the licensed operators. Additionally, this LER will be included in the required reading program for appropriate GPU Nuclear personnel.

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I MONTHLY OPERATING REPORT MAP.CH 1989 (cont'd)

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LER 39-008: LOSS OF AIR EVENT WILL PREVENT SECONDARY CONTAINMENT ISOLATION

- UUE 10 INADEQUATE SURVEILLANCE TESTING OF THE AIR SUPPORT SYSTEM As a result of NRC Generic Letter 88-14, " Instrument Air Supply System Problems Affecting Safety-Related Equipment", special testing was performed on all secondary containment isolation valves (SCIV).

The components within the scope of.the test included the check valve, accumulator, solenoid operated vent valve, actuator, and related piping. All twenty-four secondary

' containment isolation valves were determined to be in a degraded condition and

- could not be considered operable. This condition is considered reportable in accordance with 10CFR50.73(a)(2)(v)(C).

The results of the tests performed indicated that under a loss of instrument

. air event plant operators would have had insufficient time and inadequate direction to ensure secondary containment integrity remained intact.

The condition being reported is of safety significance in that secondary containment' integrity could not be assured under a loss of air event. The root cause of this occurrence is management oversight.

No formal surveillance or preventive maintenance program existed for the air support system for the SCIVs. Maintenance was performed and the tests were repeated successfully.

A formal surveillance program will' be developed.

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0841 A:04 1

I-OPERATING DATA REPORT-OPERATING STATUS 1.

DOCKET:

50-219

2. ' REPORTING PERIOD: 03/89 1

3.

UTILITY CONTACT:

' JOHN H. SEDAR, JR.

609-971-4698 4

LICENSED THERMAL POWER (MWt):

1930 5.

NAMEPLATE RATING (GROSS MWe):.

687.5 X 0.8 = 550.-

6 DESIGN ELECTRICAL RATING (NET We):_

650 j

7.

MAXIMUM DEPENDABLE CAPACITY (GROSS MWe):

642 8.

MAXIMUM DEPENDABLE CAPACITY (NET We):

620

9..IF CHANGES OCCUR AB0VE SINCE LAST REPORT, GIVE REASONS:

'NONE

10. POWER LEVEL TO WHICH RESTRICTED, IF ANY (NET We):

11. - REASON FOR RESTRICTION, IF ANY:

NONE MONTH YEAR CUHJLATIVE 12.

REPORT PERIOD HRS 744.0 2160.0 168936.0:

' 13. HOURS RX CRITICAL 63.5 63.5 106308.9 14.-

RX RESERVE SHTDWN HRS 0.0 0.0 91 8.2

15. HRS' GENERATOR ON-LINE 13.7 13.7 103557.2

16. UT RESERVE'SHTDWN HRS 0.0 0.0 1208.6

17. GROSS THERM ENER ( WH) 13550 13550 173534439

18. GROSS ELEC ENER (MWH) 2030 2030 58606214

19. NET ELEC ENER ( WH)

-3122

-10171 56249697 20.

UT SERVICE FACTOR 1.8 0.6 61.3 21.

UT AVAIL FACTOR 1.8 0.6 62.0

22. UT CAP FACTOR (MDC NET) 0.0 0.0 53.7

23. UT CAP FACTOR (DER NET) 0.0 0.0 51.2 24.

UT FORCED OUTAGE RATE 0.0 0.0 11.5

25. FORCED OUTAGE HRS 0.0 0.0 13510.7 l

26. SHUTDOWNS SCHEDULED DVER NEXT 6 MONTHS (TYPE, DATE, DURATION):

Currently shutdown for refueling outage which started September 30, 1989 27.

IF CURRENTLY SHUTDOWN ESTIM4TED STARTUP TIME: 04/22/89 1965B/0045X

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AVERAGE DAILY POWER LEVEL NET MWe DOCKET #........

50219 UNIT.......... 0YSTER CREEK #1 REPORT DATE....... April 4,1989 COMPILED BY....... JOHN H. SEDAR JR.

TELEPHONE #......

609-971-4698 MONTH MARCH,1989 DAY MW DAY MW i.

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Dyster Creek Station #1 Docket No. 50-219 REFUELING INFORMATION '- MARCH,1989 Name of Facility: Oyster Creek Station #1 Scheduled date for next refueling shutdown:

Current outage started 9/30/88 Scheduled date for restart following refueling: April 22,1989 Will refueling or resumption of operation thereafter require a Technical Specification change or. other license amendment?

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

Technical Specification Change Request No.166 was approved in November 1988.

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 Fuel 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

= 1595 (c) in dry storage 16

=

The present licensed spent fuel pool storage capacity and the size of any 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 fuel pool assuming the present licensed capacity:

Reracking of the fuel pool is in progress.

Nine (9) 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 1994 refueling outage.

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• e GPU Nuclear Corporation

" Nuclear

= r 388 Forked River, New Jersey 08731-0388 609 971-4000 Writer's Direct Dial Number:

April 14,1989 U.S. Nuclear Regulatory Commission ATTN:

Document Control Desk Washington, D.C.

20555

Subject:

Oyster Creek Nuclear Generating Station Docket No. 50-219 Monthly Operating Report In accordance with the Oyster Creek Nuclear Generating Station Operating 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 Kathy Barnes, Oyster Creek Licensing at (609)971-4390.

Very truly yours, f

E. E. Fitzpatrick Vice President and Director Oyster Creek EEF:KB:aa l

(0841 A:2)

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Enclosures j

cc:

Director Office of Inspection and Enforcement

)

U.S. Nuclear Regulatory Commission Washington, DC 20555 l

I Mr. William T. Russell, Administrator l

Region I I

U.S. Nuclear Regulatory Commission 475 Allendale Avenue King of Prussia, PA 19406 Mr. Alexander W. Dromerick, Project Manager U.S. Nuclear Regulatory Commission Division of Reactor Projects I/II Washington, DC 20555 NRC Resident Inspector Oyster Creek Nuclear Generating Station f

l GPU Nuclear Corporahon is a subsidiary of the General Pubhc Utihties Corporation j

l

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