ML20205C196

Monthly Operating Rept for Jan 1987

ML20205C196
Person / Time
Site:	Oyster Creek
Issue date:	01/31/1987
From:	Baran R, Fiedler P, Notigan D GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF RESOURCE MANAGEMENT (ORM)
References
NUDOCS 8703300132
Download: ML20205C196 (10)

Text

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MONTHLY OPERATING REPORT - JANUARY 1987 At the beginning of the report period, Oyster Creek was in a cold shutdown condition and in the third day of a planned 7-day outage. The plant was shutdown on December 29, 1986 for replacement of a faulty expansion bellows located downstream of a reheater supply line relief valve. Following required repairs, reactor startup commenced on January 6.

On January 7, power ascension was halted. The relief valve replaced during the 7-day outage was observed to be lifting. A mechanical gag was installed and load increases continued. Load increases were terminated at 338 MWe to accomodate fuel conditioning and to further evaluate corrective action concerning the lifting relief valve. The problem was subsequently corrected by adjusting the lift setpoint while the plant was operating and load increases resumed. In addition, the other relief valve on the same line was also adjusted.

While increasing load, speed control problems were experienced with motor-generator (MG) sets associated with 'C' and 'E' recirculation pumps. The scoop tubes for both MG Sets were placed in a " locked" position to maintain constant speed. On January 12, while attempting to return the 'E' recirculation pump to the " auto" mode for a power increase, the pump speed increased to maximum flow. A decision was made to remove the pump from service and continue increasing power with the remaining four (4) pumps. A plant load of 510 MWe was reached on January

13. While increasing power on January 12 and 13, 'C' recirculation pump speed was adjusted manually.

On January 14, 'B' Isolation Condenser was declared inoperable due to failure of an isolation valve to reopen after it had been closed.

Investigation revealed that the problem was attributed to a faulty spring pack in the drive portion of the valve's Limitoroue Operator. On January 16, following completion of repairs and evaluation of test data, the valve was satisfactorily tested and 'B' Isolation Condenser was returned-to-service.

On January 16, while attempting to return 'E' recirculation pump to service, a reactor scram occurred due to a high neutron flux condition which resulted from the pump's main discharge valve not being fully closed when the pump was started, although the valve did indicate closed in the control room. During scram recovery, a second scram occurred on low reactor water level which was attributed to improper mechanical pressure regulator (MPR) calibration. The MPR was subsequently calibrated and normal pressure control established.

Following the scram, the reactor was placed in a cold shutdown condition to accommodate a drywell entry for repairs to an isolation valve associated with the flydrogen/0xygen Monitoring System.

Following reouired repairs, reactor startup commenced on January 19.

During power ascension, problems developed with Intermediate Range Monitor (IRhD 17. A decision was made to interrupt the startup scouence to accommodate replacement of IRM 17.

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MON 111LY OPERATING REPORT - JANUARY 1987 %ge2 Reactor startup recommenced on January 20 and the generator placed r' on-line January 21. During power ascension, dryvell unidentified Icak ,

rate increased from 0.45 p;)m to approximately ~1.s gpm,. Investigation revealed that the leak was fran a reactor sample valve which was subseauently closed, returning unidentified ' leak rate to approximately 0.5 gpm. Reactor power and plant load. increases resumed. A maximum plant load of 660 We ws achieved on January 26.

On the evening of January 26, a water invente,ry discrepancy was' noted while transferring the contents of a Radwaste tank to the main cor. denser hotwells. Upon investigatior, it was discovered that a valve on 'the 4-inch drain line from the condensate storage tank had cracked and was leaking approximately 50 ppm into a concrete pit which diverted most of the water to a sump in the Turbine Building Basement. Approximaro1ly2000 gallons of water spilled onto the ground and froze in place. InJorder to eliminate the demand on Radwaste while efforts were underway to inlate the leak, a portable pump was installed in the pit to divert the leakage back to the condensate storage tank. ,

Although not reauired by procedure, an " Unusual Event d was declared and notifications were completed. Additionally, a plant shutdown vis r commenced while the inipact of the conditions was leing i evaluated. Once ,

it was determined that the measures taken ensured that adeouate Condensate Storage Tank inventory could be maintained, it was decided to hold power at approximately 62% (400 W e). ,

On January 27, the faulty valve was re;) laced and the " Unusual $ vent" was terminated at 1935 hours0.0224 days <br />0.538 hours <br />0.0032 weeks <br />7.362675e-4 months <br />. The ground area around the tank was subsequently recovered.

On January 30, 'A' channel of the primary containment ih drogen/0xygen Monitoring System was again declared inoparable. Tech;ical Specifications reauire repairs within 30 days. '

Plant load at the end of the report period was 652 hme, ljmited by condensate header discharge pressure.

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MONTHLY OPERATING REPORT-JANUARY 1987 q

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The following Licensee Event Reports were submitted during the month of

January 1987:

l Licensee Event Report 50-219/86-029 " Potential Inoperability of Core j 3; 3FFay/ Emergency Service Water Pump 3 Due To Inadequate Design and Procedure

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On December 6,1986, a thermal overcurrent relay for an emergency service f ~' water (ESW) punp was tested to determine if its long term current setpoint 1 was correct. Earlier inservice testing of the A ESW pump caused the pump trouble alarm to actuate. The ESW relay problem was caused by setting the

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operating current band above nameplate current to meet system flow jn ,raquroments i without considering the tolerance of the thermal relay. The

, , ,. core spray pumps, which have the same control circuitry, could exceed the

{ . relay setpoint .if operated in run-out mode. The core spray relay problem

.was caused by lack of consideration of pump operation in run-out during

,Jnitial design. The relay setpoints have been raised.

i 9 y Licensee Event Report 50-219/86-030 " Isolation Condenser "A" Isolation on j Spur 1ous H1gh F1ow 5 Mn(1";

! On December 10, 1986, the "A" isolation condenser isolated on a high flow

signal from its concensate 'r eturn line flow sensor. The high flow signal l . was caused by a small leak in the flow sensor's low pressure line. At the ,

time of the event, the plant was shutdown in a refueling outage with

reactor temperature at 163*F and pressure at atmospheric. The instrument i- lines of the; tripped flow sensor were vented and filled within two hours j

of the event. This returned the flow sensor indication to an expected shutdown value. The root cause of this event has been determined to be a i procedaval deficiency. . The surveillance procedure will be revised as necessary to correct tne deficiency. Furthermore, all surveillance 1 procHoret will be reviewed for a similar problem and revised as necessary.  :

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1 Licensee Event Report 50-219/86-031 " Reactor Building Closed Cooling Water to i

Drywell Inlation Caused by Personnel Error During Instrument Filling i A"c"hW1 tie s":

n On December 18, 1986 Reactor Building Closed Cooling Water flow to the lp ,

drywell isolated on low-low-low reactor vessel water level signals. The l signals were caused by a pressure spike in the instrument sensing line  :

1 shared by a reactor fuel ' zone level instrument which was being filled. -

l The plant was shutdown in the REFUEL mode with reactor coolant temperature 1

at 170*F. The cadse of the occurrence has been attributed to personnel

! error in not taking precautions to prevent actuation of other sensors l connected to the sensing line of the one being filled. Long term

!- corrective actions to.be taken include training of Instrument and Controls I personr.el and job planners on this event, and development of an instrument i filling and venting procedure.

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. . Lic :s-d Event Reports January 1987 t

Page 2 Licensee Event Report 50-219/86-032 " Reactor Trip on High Neutron Flux Caused By Cold Fe_e_dwater Addition Due to Operator Error":

On December 24, 1986 a reactor trip occurred due to a high neutron flux condition caused by cold feedwater addition to the reactor vessel. At the time of the event reactor power was less than 1 megawatt thermal on the Intermediate Range Neutron Monitors and being reduced to make repairs to a steam leak on the turbine first stage reheater. The cause of the event was a combination of operator errors in that personnel did not maintain an active awareness of plant system status and configuration with respect to the Feedwater Control Sy tem. Corrective action consisted of reviewing '

the event with each oncoming shift prior to their assuming startup activities. A more detailed review of the event stressing the need to maintain an active awareness of plant status and conditions will bei incorporated in operator training, once the final transient assessment report is complete. N Licensee Event Report 50-219/86-033 " Standby Gas Treatment Initiation Caused by Ground on ARM Ribbon Cable Due to Personnel Error":

On December 21, 1986 Reactor Building Ventilation was isolated and the Standby Gas Treatment System was initiated when an Area Radiation Monitor trip unit was inadvertently grounded during maintenance. At the time of the event, the plant was in the startup mode with the reactor subcri tical . An instrument technician caused the ground by pinching a ribbon cable when inserting one of the monitor trip units. The root causes of this event were the personnel errors of installing the ribbon cable so that it was susceptible to pinching and not taking sufficient care when inserting the ARM trip unit. Corrective actions were taken to replace, the pinched ribbon cable and the shorted out power supply.

Additionally, maintenance personnel will be instructed on the caution '

required when working on equipment.

Licensee Event Report 50-219/86-034 " Manual Scram Due to Inability to Maintain Condenser Vacuum Caused by Equipment Failure": ,

On December 29, 1986, while operating at 38.8 percent power, a steam leak ,

was detected on an expansion joint connecting discharge piping of a lifted steam reheater relief valve to the "B" condenser. An unsuccessful attempt was made to gag the lifted relief valve while the second stage reheaters were out of service to allow patching of the joint. Reactor power was reduced and. the turbine was secured to reduce steam pressure at the joint. Air entered the condenser and vacuum began to decrease. Condenser vacuum stabilized at 23.8 inches Hg. It was determined that further attempts to repai r the leak would be unsuccessful and that condenser vacuum could not be maintained above the scram setpoint during a normal plant shutdown. The operators reviewed the procedures and manually scrammed the reactor as a controlled evol ution. Subsequent review of saturation temperature readings recorded by the computer showed that a cooldown of 101 *F over a one hour period had occurred following the scram. The Technical Specification ifmit is 100*F in a one hour period.

The cause of the loss of vacuum was the expansion joint failure.

Inadequate procedural guidance was the cause of the high cooldown rate.

The procedure limit was reduced to prevent a recurrence. i

. . Licens:e Event Reports January 1987 Pag) 3 Licensee Event Report 50-219/86-035 " Containment Penetration Found Degraded Due to Isolation Valves Actuator / Valve Linkages Out of Adjustment":

After experiencing problems venting the torus following a shutdown, a diagnostic leak test was performed on two containment isolation valves.

The results of the leak test performed on December 31, 1986, revealed that both valves were leaking excessively. Maintenance personnel rebuilt and adjusted the valve / actuator linkage on one valve and adjusted the stem / actuator coupling on the other valve. Following maintenance a local leak rate test was performed and test results were acceptable. The main contributor to the leakage experienced was the linkage between the actuator and valve being out of adjustment. Since the two valves involved in this occurrence are in series in the same penetration, this situation could have prevented the fulfillment of the safety function of the primary containment, that is to control the release of radioactive material and/or mitigate the consequences of an accident. Due to the seriousness of this occurrence, coupled with the fact that the root cause has not been identified, an independent root cause investigation will be performed. A followup LER will be forwarded when the investigation is complete.

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. .. OPERATING DATA REPORT OPERATING STA'IUS

1. DOCKET: 50-219

2. REPORTING PERIOD: JANUARY, 1987

3. UTILITY 00tfrACT: DONALD V. NOTIGAN 609-971-4695

4. LICENSED THERMAL POWER (MWt): 1930

5. NAMEPLATE RATING-(GROSS MWe): 687.5 X 0.8 = 550

6. DESIGN ELECTRICAL RATING (NET MWe): 650

7. MAXIMUM DEPENDABLE CAPACITY (GROSS MWe): 650

8. MAXIMUM DEPENDABLE CAPACITY (NET MWe): 620

9. IF CHANGES OCCUR ABOVE SINCE LAST REPORT, GIVE REASONS: NONE

10. POWER LEVEL 'IO WHICH RESTRICIED, IF ANY (NET MWe): N/A

11. REASON FOR RESTRICTION, IF ANY: NONE MONTH YEAR CUMULATIVE

12. REPORT PERIOD HRS 744.0 744.0 149977.0

13. HOURS RX CRITICAL 536.3 536.3 95372.8

14. RX RESERVE SHTDWN HRS 0.0 0.0 918.2

15. HRS GENERA'IOR ON-LINE 479.2 479.2 92849.2

16. Ur RESERVE SHTDWN HRS 0.0 0.0 1208.6

17. GROSS THERM ENER (MWH) 653104 653104 153609489

18. GROSS ELEC ENER (MWH) 214860 214860 51883105

19. NET ELEC ENER (MWH) 203424 203424 49813501

20. Ur SERVICE FACIOR 64.4 64.4 61.9

21. UT AVAIL FACIOR 64.4 64.4 62.7

22. UT CAP FACIOR (MDC NET) 44.1 44.1 53.6

23. UT CAP FACIOR (DER NET) 42.1 42.1 51.1

24. Ur FORCED OUTAGE RATE 35.6 35.6 10.5

25. FORCED OUTAGE HRS 264.8 264.8 10916.6

26. SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, DURATION): N/A

27. IF CURRENTLY SHUIDOWN ESTIMATED STARTUP TIME: N/A 1965B ,

,b AVERAGE DAILY POWER LEVEL ,

NET MWe DOCKET f. . . . . . . . 50-219 I

UNIT. . . . . . . . . . . OYSTER CREEK #1 REPORT DATE . . . . . . . FEBRUARY 6, 1987 COMPILED BY . . . . . . . DONALD V. NOTIGAN TELEPHONE i . . . . . . 609-971-4695 MONTH JANUARY, 1987 DAY MW DAY MW

1. 0 16. 401

2. 0 17. 0 3.- 0 18. 0

4. 0 19. 0

5. 0 20. 0

6. 0 21. 41

7. 151 22, 219

8. 310 23. 473

9. 291 24. 568 10, 288 25. 607 11, 290 26. 630

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12. 329 27. 386

13. 482 28. 432 l

14, 471 29, 482

15. 458 30. 621 31, 629 19688 l

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

Scheduled date for restart following refueling: N/A Will refueling or resumption'of operation thereafter require a Technical Specification change or other license amendment?

No Scheduled date(s) for submitting proposed licensing action and supporting information:

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. New operating procedures, if necessary, will be submitted at a later date.

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, I-in number of fuel assemblies:

present licensed capacity: 2600 l The projected date of the last refueling that can be discharged to the spent

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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 1990 refueling outage.

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50-219 UNITSilufDOWNS AND POWER REDUCTIONS UNITNAME Ovster Creek

• January 1987 DATE COMPLETED gy R. Baran REPORT MONill .Tn mm v Y TEl.Eril0NE 971-4640 u o E 5g .,.E5

- "g Licensee Ee, , Cause & Corrective No. Date g 3g  ;; .3 s 5 Event gi {*3 go Action Io F E 3;g g Repost # us u , Prevent Recussence fE 4 o

O 48 12/29/86 F 201.6 A 1 N/A N/A N/A Manual scram during shutdown process due to vacuum decreasing ,

Required repairs of the cross-around relief valve expansion

' joints.

59 1/16/87 F 115.1 A 3 N/A N/A N/A Reactor scram on high neutron fl' tx 1 when "E" reactor recirc. pump wa: ;

started *with its discharge valve not fully closed thus inserting posi- :ive reactivity.

50 1/26/87 N/A N/A A N/A N/A N/A N/A Approximately 39% reduction in generator output (660 MWe to 404 MWe) due to a leak in the condensate storage tank drain pip'e.

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1 2 3 4 I F: Fosced Reason: Method: Exhibit G.lastsuctions o 5: Scheduled A-Equipment Failure (Esplain) lunual for Pseparatica of Data -

B-Maintenance of Test 2Mnual Scsasa. Entry Sheets for Licensee C-Refueling 3-Automalle Sesam. Event Repost(LER) File (NUREG.

D-Regulatosy Restriction 4-Oshes (Explain) 0161)

E Operatos Tsaining & Ucense Exandnation F-Administrative . 5 .

G Operational Es;os (Emplain) Exhibli! SaaneSousee 19117) 11 Oth:r (Esplai 0  !

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l GPU Nuclear Corporation NggIgf Post Office Box 388 Route 9 South Forked River New Jersey 08731-0388 609 971-4000 Writer's Direct Dial Number:

February 17, 1987 Director Office of Management Information U.S. Nuclear Regulatory Commission Washington, DC 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 Mr. Joseph D. Kowalski, Oyster Creek Licensing Manager at (609)971-4643.

Very truly yours, f

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Tiedler An

Yice President and Director Oyster Creek PBF:KB: dam (0841 A)

Enclosures cc: Director (10)

Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, DC 20555 Dr. Thomas E. Murley, Administrator Region I U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406 Mr. Jack N. Donohew, Jr.

U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue, Phillips Bldg.

Bethesda, MD 20014 NRC Resident Ins ector Oyster Creek Nuc ear Generating Station 2424 GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation

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