ML20238D882

Monthly Operating Rept for Nov 1987

ML20238D882
Person / Time
Site:	Oyster Creek
Issue date:	11/30/1987
From:	Baran R, Fiedler P, Sedar J GENERAL PUBLIC UTILITIES CORP.
To:	NRC
References
NUDOCS 8801040493
Download: ML20238D882 (9)

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MONTHLY OPERATING REPORT NOVEMBER 1987 4

The following Licensee Event Reports were submitted during the month of November 1987:

LER 87-036 "High Radiation Area Technical Specification Violation Due to Personnel Error and Procedural Non-Compliance" l

On October 13, 1987, with the reactor in a cold shutdown condition and the drywell open for access, four (4) workers entered a locked high radiation area  ;

without the minimum radiation monitoring equipment required by procedure and station technical specifications. In addition, the locked radiation door to the area was not properly controlled as required by procedure' and station technical specifications. Of the workers involved (four (4) contractor j personnel) three (3) had entered the area on the previous day meeting all

• requi rements. The day of the violation each man erroneously assumed one of them obtained the required alarming dosimeter. They were aware of the requi rement. The cause of this event is attributed to personnel error. The.

failure to properly control the entry point is attributed to poor Radiological Controls practice when controlling dryMell access. Control of the entry point was informally being passed to another individual. This caused a breakdown of positive control and was not being performed per station procedure. The

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personnel who entered the area performed their task and received doses below the expected levels. Three (3) of the contractors involved were terminated.

The fourth individual received a written reprimand, as he had not been involved on the previous day and was not as familiar with the job as the other i ndivi dual s. Their site coordinator has prepared a lesson plan for current and future workers to ensure a better understanding of site controls.

Radiological Controls issued strict guidance to field personnel to control drywell access in accordance with station procedures, LER 87-037 " Technical Specification Violation Caused by Failure to Perform Two i Monthly Dose Calculations Due to Personnel Error" '

On October 9,1987 it was discovered that the monthly calculations of dose due to radioiodine and particulate in gaseous effluents were not performed for July and August 1987. These calculations are required by the plant's Radiological Effluent Technical Specifications. At the time of discovery the plant was shut down for maintenance, but the plant had been operating at full power for most of July and August. The occurrence is attributed to personnel j error due to inadequate training. The safety significance of this event is i minimal since data was regularly collected, no excursions were noted, and 1 results of the subsequent calculations showed releases were well below the '

annual limit. Immediate corrective action was taken to complete the calculations upon discovery of the condition. Future corrective action is training of more personnel on performance of dose calculations.

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Licensee Event Reports

. November 1987 Page 2 - -

LER 87-039 " Appendix R Criteria Not Met Due to Existing Plant Drawing Deficiency, Discovered During Modification Document Update" On Octolfer 14, 1987, it was discovered that a signal circuit from a lockout relay, which prevents the Emergency Diesel Generator #2 (EDG-2) breaker from closing in the event of a 4160 volt electrical bus fault, does not comply with 10CFR50, Appendix R. 1he circuit is not protected from a fire in a fire zone for which response relies upon the operability of EDG-2. The ability to close the EDG-2 breaker could be lost in an Appendix R fire event without an actual electrical bus fault. The plant was shutdown at the time of discovery, however, the condition has existed since startup from the December 1986 refueling outage. The cause of the condition is a design deficiency. The root cause of the deficiency was a discrepancy between actual circuitry in the field and that which was reflected in an existing plant drawing used for the design of the modification. It is significant in that it could have prevented a safe shutdown in response to a fire event. Corrective action has been taken to i modify the circuitry to comply with 10CFR50 Appendix R criteria.

LER 87-040 " Torus 'to Oxygen Sample Line Doesn't Meet Single Failure Criteria j Due to Design Deficiency" The plant's torus oxygen sample line isolation valves' logic does not meet single failure criteria as required by NRC general design criteria. This condition was determined reportable on October 16, 1987, while the plant was )

shut down for maintenance. The apparent cause of the occurrence is a design ]

deficiency which has been present since initial plant startup. This condition is significant in that the torus oxygen sample line could fail to close due to a single relay failure which could place the plant outside its design basis containment leak rate during an accident. A conservative analysis indicates that, under the design basis loss of coolant accident with this single failure present, operators would have to diagnose and take corrective action within four hours to prevent exceeding 10 CFR 100 limits. Corrective action will be taken to modify the valve isolation circuit to meet single failure criteria in the next refueling outage. Until the modification is installed, direction by means of a standing order and procedural instructions will be provided to operators to close these valves upon a containment isolation.

LER 87-042, " Potential Inoperability of Standby Gas Treatment System Due to I Design Error" l On October 26, 1987 it was determined that a condition which could prevent the fulfillment of the safety function of the Standby Gas Treatinent System (SGTS) i exists. The plant's drywell purge supply valves do not close on a reactor building isolation signal. A secondary containment in-leakage path is created when the drywell is open or being ventilated and reactor building isolation occurs with the SGTS initiated. Under these conditions the SGTS may not be able to maintain the reactor building vacuum required by Technical Specifications. At the time of discovery, the plant was shut down for  ;

maintenance; however, this condition has existed since initial plant

Licensee Event Reports a

November 1987 Page 3 -

operation. The apparent cause of the condition is a design deficiency. The conditish is significant in that a system needed to control the release of radioactive material could be prevented from' fulfilling its safety function.

Immediate corrective action was taken to secure the valves in the closed position. ~ A modification was installed to provide a reactor building isolation signal to two of' the four drywell purge supply valves.

LER 87-044 " System Systems Unable to Function on Loss of Offsite Power During Battery Maintenance Activities Due to Personnel Error" For the period October 13, 1987 through October 19, 1987 Emergency Diesel Generator #2 and the 'C' main station battery, which supplies control power to emergency diesel generator (EDG) #1, Division A 4160 volt and 460 volt switchgear, were both out of service. Redundant trains of safety related equipment required to be operable would not have functioned as designed upon a loss of offsite power during this period. The condition was discovered on October 30, 1987 dnd occurred during a maintenance outage. The root cause of the event is personnel error. The core spray and standby gas treatment systems, which are required to be operable while shut down, would not have automatically initiated during an accident with a loss of offsite power. The significance of this is minimized by: the presence of a _ fire water system to provide mskeup water to the reactor; the fact that the evolutions which caused the event are only performed when in cold shutdown; and the existence of DC Power Diagnostic and Restoration procedures. Corrective actions include procedure revisions and required reading.

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hKNiliLY OPERATING REPORT - MNEMBFR 1987 At the beginning of the report period, Oyster Creek was shutdown with the  ;

E1U6/11M Outage in progress.  !

On November 6, the NRC approved plant restart following an investigation of the Safety Limit Violation that occurred on September 11.

On h ovember 18, a strike vote was approved by Union personnel.

Supervisory personnel assumed all licensed duties and preparations for plant startup continued.

Plant startup commenced at 2:45 p.m. on November 20. Reactor power was stabilized below criticality due to a faulty 102 valve (drive-withdrawal isolation valve) on CRD hydraulic control unit 30-43. The valve was repaired and startup was resumed at 7:00 a.m. on November 22. At 8:26 a.m. the reactor was brought to criticality.

A drywell inspection at 1000 psig reactor pressure was satisfactorily completed and primary containment inerting commenced on November 23 at 3:00 a.m. The reactor mode switch was placed in "RUN" at 2:23 p.m. and containment inerting,was completed at 11:50 p.m. on the same day.

On November .23 at 2:52 p.m. while placing the generator on-line, a generator / turbine trip occurred due to actuation of the generator ground backup relay 59B. A subsequent investigation determined the relay actuation was due to a malfunction of a potential transformer drawer interlock contact. Power was increased and the generator was placed on-line at 5:35 a.m. on November 24. By the end of the day, power was increased to 71.6% thermal power (450 We).

Power was increased per Core Group direction and by the end of the report period, thermal power was 1927 We and generator load was 670 We.

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Oyster Creek Station #1 Docket No. 50-219-REFUELING INFORMATION - NOVEMBER, 1987

1. v Name of Facility: Oyster Creek Station #1 Scheduled date for next refueling shutdown: N/A f Scheduled date for restart following refueling:

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

Yes k

Scheduled date(s) for submitting proposed licensing action and supporting- l infomation: .

l March 31, 1988

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1 Important licensing considerations associated with refueling, e.g. , new or -

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

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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 l l there any anticipated.

The number of fuel assemblies (a) in the core = -560 (b) in the spent fuel storage pool = 1392 l (c) in dry storage = 20 1 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 1

The projected date of the last refueling that can be discharged to the spent 1 I

fuel pool assuming the present licensed capacity:

Reracking of the fuel pool is in progress. Nine (9) out of ten (10)'

racks have been astalled to date. When reracking is completed, i discharge capacity to the spent fuel pool will be available until 1994 refueling outage. l

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AVERAGE DAILY POWER LEVE1 NET hMe DOCKET #. . . . . . . . 5 0-219 UNIT. . . . . . . . . . 0YSTER CREEK #1 REPORT DATE . . . . . . .DECIMBER 02, 1987 COMPILED BY . . . . . . . JOIN H. SEDAR, JR.

TELEPii0NE # . . . . . . 609-971-4698 i

MONDi NOVEMBER, 1987 DAY FM DAY FM

1. 0 16. O I

2. 0 17. 0

3. 0 18. O i I

4. 0 19. O I i

5. 0 20. 0

6. 0 21. 0 l

7. 0 22. O i I

8. 0 23. 0

9. 0 24. 209

10. 0 25. 536

11. 0 26. 597

12. 0 27. 644

13. 0 28. 647

14. 0 29, 648

15. 0 30. 648 1968B

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. OPERATING DATA REPORT

. OPERATING STATUS l 1. DOCKET: 50-219

2. REPORTING PERIOD: NOVEMBER, 1987

3. (Tr4ITY CONTACT: JOHN H. SEDAR, JR. 609-971-4698

4. LICENSED THERMAL POWER (MWt): 1930

5. NAMEPLATE RATING (GROSS We): 687.5 X 0.8 = 550 l

6. DESIGN ELEC'IRICAL RATING (NET We): 650

7. MAXIMUM DEPENDABLE CAPACITY (GROSS MWe): 650

8. MAXIMUM DEPENDABLE CAPACITY (NET We): 620

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

10. POWER LEVEL 'ID WHICH RESTRICTED, IF ANY (NET We): 0 from 11/1 to 11/6 N/A after 11/6

11. REASON FOR RESTRICTION, IF ANY: NRC imposed shutdown from 11/1 to 11/6.

j MONTH YEAR CIMJLATIVE 1

12. REPORT PERIOD 1RS 720.0 8016.0 157249.0 )

13. HOURS RX CRITICAL 207.6 4875.9 99712.4

14. RX RESERVE SHTDWN IRS 0. 0 0.0 1208.6 )

15. HRS GENERATOR ON-LINE 162.4 4678.9 97048.9' 1 16. ITT RESERVE SHTDWN 1RS 0.0 0.0 1208.6

17. GROSS THERM ENER (MWH) 297800 8260404 161216789

18. 97940 2752540 54420785  ;

GROSS ELEC ENER (WH) 1 I

19. NET ELEC ENER (WH) 91280 2630859 '52240936

20. UT SERVICE FAC'IOR 22.6 58.4 61.7

21. tTT AVAIL FACTOR 22.6 58.4 62.5 I

22. UT CAP FACTOR (MDC NET) 20.4 52.9 53.6

23. ITT CAP FACIDR (DER NET) 19.5 50.5 51. 1

24. [TT FORCED OUTAGE RATE 0.0 30.3 11.6

25. FORCED OUTAGE IRS 0. 0 2034.5 12686.3

26. SilUTDOWNS SCHEDUIID OVER NEXT 6 MON'IliS (TYPE, DATE, DURATION):

27. IF CURRENTLY SHUTDOWN ESTIMATED STARTUP TIME: N/A i

GPU Nuclear Corporation

', h.p g gf Post Office Box 388 Route 9 South Forked River.New Jersey 087310388 4 609 971-4000 Writer's Direct Dial Number:

Director Office of Management Information December 15, 1987 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, na L' g n p@[g[6h- '

Vice President and Director Oyster Creek 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 g O> ,/

King of Prussia, PA 19406 4 -

Mr. Alexander W. Dromerick, Project Manager @ -

U.S. Nuclear Regulatory Comission /y Division of Reactor Projects I/II 7920 Norfolk Avenue, Phillips Bldg. t O',0 r

• g Bethesda, MD 20014 i

NRC Resident Inspector Oyster Creek Nuclear Generating Station GPU Nuclear Corporat:On is a subsidiary of the General Pubhc Utihties Corporation

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