ML20236W111

Monthly Operating Rept for Sept 1987

ML20236W111
Person / Time
Site:	Oyster Creek
Issue date:	09/30/1987
From:	Fiedler P GENERAL PUBLIC UTILITIES CORP.
To:	NRC
References
NUDOCS 8712070282
Download: ML20236W111 (10)

Text

NRC Monthly Operating Report - September 1987 At the oeginning of the report period, Oyster Creek operated at 99", core t h > r::u l powe r wi t h a gene ra t o r l oad o f 044 h ; 4 MWe.

1 on September o, plant load wts reduce 1 to 3no % to repai r No. 2 main I

flash tank manway.

Atter further evaluat ion, it was determined that the flash tank leak cculd not safely be repaired while on-line.

At 9:00 p.m.

on September 9, a shutdown was begun.

The generator was taken off-line at 1:07 a.m.

on September 10, and cold shutdown was achieved later that day.

On September 11, a safety limit was violated while operating the recirculation pump discharge valves.

The violation occurred while attempting to secure the recirculation pumps in response to leakage from Reactor Building Closed Cooling Water drywell isolation valve V-5-167.

An investigation is being conducted into the safety limit violation.

The plant will reinain shut down until NRC approval to start up is granted.

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NRC Monthly Operating Report - September 1987 At toe beginning of the report pe r i od, (:.ws t e r Creek ope ra M! a t 99', core theraal power with a generatur load of o!! 'f,e ;.1 %.

plant lo ui., o reduced t>

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' 3. - to r 7. t t, c so.

n,:

On Jepte:

m flash tank :anway.

M ter further evaluation, it was determinH that toe t' lash tank leal could not sately be repaired win t e on-line.

At 9:00 p.m.

on September 9, a shutdown was begun.

The generator was taken of f-line. at 1:07 a.m.

on September 10, and cold shutdown was achieved later that day.

On September 11, a safety limit was violated while operating the recirculation pump discharge valves.

The violation occurred while attempting to secure the recirculation pumps in response to leakage from Reactor Building Closed Cooling Water drywell isolation valve V-5-167.

An investigation is being conducted into the safety limit violation.

The plant will remain shut down until NRC approval to start up is granted.

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MONTHLY _0PERATING REPORT SEPTEMBER 1987 Ine foll ming Licensee Event leparts cre subal tted during tw..wnt ) of September 1937:

LER 87-028 " Main Steam Isolation Valve Closure Caused by Design Deficiency During Surveillance Test":

On August 2,1987 at 0340 hours0.00394 days <br />0.0944 hours <br />5.621693e-4 weeks <br />1.2937e-4 months <br /> a main steam isolation valve (MSIV) closure was inadvertently initiated during an intermediate range neutron monitor (IRli) front panel surveillance.

At the time the plant was in the REFUEL' mode with reactor coolant temperature at.130*F, To test the operation of the IRM range 10 interlock with the main steam line low pressure MSIV automatic closure signal, one channel of the Reactor Protection System (RPS) is jumpered to prevent MSIY closure while a trip-signal is introduced.

When technicians installed the reqdired jumpers and tested.them for tightness, one. end of one -

jumper fell off and shorted to ground, blowing an RPS fuse.

This caused an MSIV. isolation signal in RPS channel 2.

When another isolation signal was intentionally introduced later in the surveillance an MSIV closure occurred.

Investigation uncovered the blown fuse, which was replaced, and the surveillance was completed at 0418 hours0.00484 days <br />0.116 hours <br />6.911376e-4 weeks <br />1.59049e-4 months <br />.

The root cause of the event is that the circuit being tested is not configured to facilitate testing.

Contributing causes were' the jumpering technique of the technician and failure of the Control Room operators to. fully investigate a suspected actuation.

This evolution is not performed during power operations and an MSIV closure while shut down has minimal effect on the reactor, therefore the safety i

-significance of this event is minimal.

This report will be made required l

reading for instrument technicians, electricians, plant engineering department engineers and operators. Modifications to facilitate testing are being evaluated.

LER 87-029 "High Reactor Pressure Scram Due to Air Leak From Dislodged Air Test Pilot Valve Caused by Incorrect Mounting Cap Screw Length":

On July 30,1987 at 0500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> a high reactor pressure scram occurred when an air leak caused a Main Steam Isolation Valve (MSIV) to close with reactor power. greater than the level where one main steam line alone is sufficient to maintain reactor pressure below the scram setpoint.

The plant was operating at approximately 97% rated power prior to this event with reactor pressure at 1020 psig and reactor temperature at 524'F.

During a 5% closure test of MSIV NS04A, the air test pilot valve partially dislodged from its mounting causing the MSIV to close beyond 5% and stabilize at an intermediate position.

Thi s l

produced a low control air pressure alarm.

The operators quickly diagnosed l

the problem and began reducing power and started two additional air compressors. Fifteen minutes had elapsed and reactor power had been reduced l

Licensee Event Reports September 1987 Page 2 to approximately 70; when the air test pilot valve di slodged completely

!osing the f tSIV.

This caused a high reactor pressure scram.

The cause of t,ts event i s the Ji r test pil n vil ve 1]unt 0; :3p sce m aero 1.'1 inch 4

soorter tiu1 vendor speci fications.

Tne 11slvjyrj dir te n pilat el ve aM 311 >tner 'GIV ai r pil]t v 31 ves m _ - > inn

' ai N ar,- leojtn :p sc rws.

Tne :GIl air pilat valve iutd iniw, exem

.n l l ae rm s a t,

i den ti ff the ;; rope-length screc ',c

" r. x e 12 m u n;e.37-030 " Lightning Arrestor Insulator Failure Induced Voltage Transient Caused Containment Isolation and SBGTS Initiation Due to Automatic Bus Transfer Time f ;

Exceeding RPS Relay Dropout Time":

On April 22, 1986 at approximately 0700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br />, primary and secondary a

containments isolated and the Standby Gas Treatment System (SBGTS) initiated g

l as a result of a voltage transient caused by a lightning arrestor insulator failure.

The voltage transient caused Vital AC Power Panel 1 (VACP-1) to transfer to its alternate power supply. The power supply transfer caused

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several Reactor Protection System (RPS) relays to deenergize, causing the

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containment isolat' ions and SBGTS initiation.

At the time of this event, the reactor mode switch was locked in the SHUTDOWN position, the vessel head had

/

been removed and the vessel and reactor cavity had been flooded in preparation for refueling.

The shortest transfer time achievable with the automatic transfer switch exceeds the dropout time for the RPS relays.

The isolation signal was reset and SBGTS was secured following the event.

The safety significance ef this event is minimal as only a challenge to containment isolation and SBGTS initiation logic circuits occurred.

To prevent recurrence of this event, engineering has proposed a continuous power supply be connected to the circuit which contains the relays that cause coritainment isolations and SBGTS initiations when certain power interruptions occur.

This proposal is currently being considered by corporate management for approval.

LER 87-031 " Violation of High Radiation Area Technical Specifications Caused by Personnel Error During Response to Fire Alarm":

On August 27,1987 at 2250 hours0.026 days <br />0.625 hours <br />0.00372 weeks <br />8.56125e-4 months <br /> plant personnel violated high radiation area technical specifications and procedures. At the time, the plant was operating at 99% power. At 2239 hours0.0259 days <br />0.622 hours <br />0.0037 weeks <br />8.519395e-4 months <br /> a main transformer / turbine area fire alarm was received in the control room.

Operators investigated and found no fire at the main transformer.

At 2242 hours0.0259 days <br />0.623 hours <br />0.00371 weeks <br />8.53081e-4 months <br /> a moisture separator high water level alarm was received and a supervisor (senior reactor operator licensed) joined an investigating control room operator at the condenser bay entrance.

The sprinkler system to the condenser bay was manually isolated because the supervisor suspected that the fire water was spraying on instrumentation and causing it to behave erratically, threatening a turbine trip.

The supervisor then directed the control room operator to climb over the locked high radiation area door and open it from the inside before the requested radiological controls technician arrived.

The supervisor believed the threat of a possible fire condition outweighed the high radiation area access

Licensee Event Reports September 1987 Page 3 t

requirements.

Toa c use of the event was personnel errar, an1 th3 saf?ty si jni fic ance was jeteraiac j to tw ainimai.

No overexposure af personnel ces>lted.

Curce: tie 3: 1: 0; in:lude oMr3 tor training, pron sion of 001i tJring eqJipnen; tv op ~;;100; personnel for emergency usa, and review of

• ais ? vent with all op's t ~ ; 1r ! 9 i sup9-visor s.

LER 87-032 "A0G H2 Analyzer Not Calibrated in Accordance with Tech Spec Requirements Due to Inadequate Review of RETS Amendment":

On Angust 24, 1987 at approximately ICIS hours, it was discovered that the Augmented Offgas System H2 analyzer had been calibrated using a standard gas sample of a known volume of H2 in air rather than in N2 as required by Technical Specifications (TS).

The plant hcd been operating in various power levels and modes during the time period the H2 analyzer was r.ot calibrated in accordance with TS requirements.

This quarterly TS requirement was part of the Radiological Effluent Technical Specifications (RETS) amendment which became effective November 20, 1986.

On September 5, 1987, the A0G system H2 analyzer was calibrated in accordance with TS requirements using a newly developed procedure.

The root cause of this event was personnel error in that the plant engineering review of the RETS amendment did not identify the RETS inconsistency with the existing H2 analyzer calibration method nor did they initially identify the need for a procedure to be developed for this calibration.

For corrective action this LER will be required reading for all involved departments to stress the importance of a thorough review of plant procedures and practices for compliance with TS changes.

Since this event, Oyster Creek Licensing management has altered its policy and is issuing Licensing Actions Items for tracking requisite procedure changes resulting from TS changes, dmd:0841 A

Oyster Creek Station til Docket No. 50-214 iWPtONG INFORM \\ TION - SEi'fP1RE1 IW N:ina of Facility: ovstor Croek St uion lil Scheduled date for next refueling shutdowm N'\\

Scheduled date for restart following refueling:

Will refueling or resumption of operation thereafter reqiire a Technical Specification change or other license amendment?

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

March 31, 1988 Important licensing considerations associated with refueling, e.g.,

new or different fuel design or supplier, unreviewed design or performake 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 = 1392 (c) in dry storage 20

=

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.

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 1994 refueling outage.

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OPERATING DATA REPORT l

OPERATING STATUS l

'1.

DOCKET:

'i0-219 2.

REPORTING PERIOD: SEPTIMBER, 1987 3.

UTILITY CONTACT:

.MN f L SED \\R, JR.

60?-9~1-In H J.

LlCfNSED T:tEIOt\\L POER L ' tit ):

1Ma 5.

NNIEPLATE RATING (GR9SS SMe):

087.I X 0.8 = 350 6

DESIGN ELECTRICAL RATING (NEF $Me):

650 1

7.

MAXIMUM DEPENDABLE CAPACITY (GROSS hMe):

650 8.

MAXIMUM DEPENDABLE CAPACITY (NET hMe):

620 9.

IF CHANGES OCCUR ABOVE SINCE LAST REPORT, GIVE REASONS:

NONE 10.

POWER LEVEL To hBICH RESTRICTED, IF ANY (NET NMe):

0

11. REASON FOR RESTRICTION, IF ANY: NRC IMPOSED SilUTD0hN MONTH YEAR

, CUMULATIVE

12. REPORT PERIOD HRS 720.0 6551.0 155784.0

13. HOURS RX CRITICAL 219.5 4668.3 99504.8 14.

RX RESERVE SHTDhN HRS 0.0 0.0 918.2

15. HRS GENERATOR ON-LINE 217.1 4516.5 96886.5

16. [JT RESERVE SHTDh'N HRS 0.0 0.0 1208.6 17 GROSS THERM ENER (SMH) 361300 7962604 160918989

)8.

GROSS ELEC ENER (hMI) 118150 2654600 54322845

19. NET ELEC ENER (MhH) 111074 2542045 52152122

20. [TT SERVICE FACTOR 30.2 68.9 62.2

21. [TT AVAIL FACTOR 30.2 68.9 63.0 22.

UT CAP FACTOR (MDC NET) 24.9 62.6 54.0

23. (TT CAP FACTOR (DER NET) 23.7 59.7 51.5

24. UT FORCED OUTAGE RATE 69.8 31.1 11.6 25.

FORCED OUTAGE HRS 502.9 2034.5 12686.3 26.

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

N/A 27.

IF CURRFNftY SHUTD0hN ESTIMATED STARTUP TIME:

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AVERAGE DAILY POWER LEVEL NET %

DOCKET #.

.50-219 (IM T.

0YSTER CREEK #l REPORT DATE....

.0TOBER 0 5, 1937 COMPILED BY,

. JOHN H. SEDAR, JR.

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GPU Nuclear Corporation Sz U NUCIBar PSt=

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Octo be r 16, 1987 Of fice of :.ianagement in f ormati on U.S. Nuclear Regulatory Commission Washington, DC 20555 I

Dear Sir:

Suoject:

Oyster Creek Nuclear Generating Station Docket No. 50-219 Monthly Operating Report In accordance with the Oyster Creek Nuclear Generating Station Operating License No. OPR-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.

-Mytr.ly;1 burs,

[7

/W PterB.Fijler ice Presid t and Director Oyster Cree (

PBF:KB:dmd(0841 A)

Enclosures cc: Director (10)

Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, DC 20555 Mr. William T. Russell, Administrator Region I U.S. Nuclear Regulatory Commission 631 Park Avenue

// /

King of Prussia, PA 19406 I

i Mr. Alexander W. Dromerick, Project Manager U.S. Nuclear Regulatory Commission j

Division of Reactor Projects I/II pgsicupT o CRIG W 7920 Norfolk Avenue, Phillips Bldg.

f

, g certified M #

Bethesda, MD 20014 NRC Resident Inspector r

Oyster Creek Nuclear Generating Station

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k GPU Nuclear Corporation is a subsia:ary of the Generw Pubhc Utmties Corporaoon

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