ML20237K787
| ML20237K787 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 06/30/1987 |
| From: | Baran R, Fiedler P GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8708190405 | |
| Download: ML20237K787 (8) | |
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MONTHLY OPERATING REPORT JUNE 1987 The following Licensee Event Reports were submitted during the month of June 1987:
Licensee Event Report 50-219/87-024 - " Failure to Post a Fire Watch for a Non-Functional Fi re Barrier Due to Personnel Error in Failing to Follow Procedure" On April 26, 1987 during a plant outage, "thermo-lag" fire barrier material installed during the 10 CFR 50, Appendix R modification work was removed to facilitate maintenance on drywell penetration number 54.
A fire watch ~ patrol was posted approximately 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> after the fire barrier was removed, exceeding Technical Specifications requirements _for a fire watch to be posted within one hour.
The plant was in a cold shutdown condition with the reactor coolant temperature less than 212*F.
The cause of the event was attributed to personnel error in failure to follow established procedures governing work practice.
The safety significance of the event is minimal because the plant was in a cold shutdown condition.
Immediate corrective action was taken to ensure a fire watch was posted.
Future corrective actions include job supervisor
- training, required
- reading, and improved instructions to contractors.
Licensee Event Report 50-219/87-025
" Primary Containment Vent and Purge Valves had Maximum Stroke in Excess of Design Limit Due to Installation Procedure Inadequacy" On May 7, 1987 four of ten primary containment vent and purge, valves were 1
found to have a maximum opening stroke greater than 30*.
This is contrary to a commitment Oyster Creek made to the Nuclear Regulatory Commission (NRC).
The plant was in cold shutdown at the time of the discovery, but had been operating through several cycles since the original installation of the valve I
restrictors in 1980.
The NRC had required the containment vent and purge valves be restricted to assure these valves would close during a loss of Coolant Accident (LOCA) in the primary containment.
The Oyster Creek analysis i
specified a 30' open restriction was necessary to assure closure.
A preliminary re-analysis indicates the valves would close from 35'-40* open during a LOCA inside containment.
Since the valve found with the greatest opening stroke opened to 35*, all the containment vent and purge valves would i
probably have closed under LOCA conditions.
The cause of this event is procedural inadequacy in that the valve restrictor installation procedure did not speci fy an accurate method of measuring valve position.
The underrestricted valves were immediately adjusted after discovery to meet commitment limits.
The reanalysis will be formalized and the results will be included in a supplemental LER.
Instructions will be written to specify a new accurate valve position measuring method.
These instructions will be incorporated into the appropriate vendor manuals.
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- Licensee Event Reports i
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June 1987 Page 2 L
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Licensee Event Report 50-219/87-026
" Temporary Variations Found Unacceptable Due to Inadequate Safety Reviews" As a result of Licensee Event Report 87-021, a special review committee was established to review all temporary variations in effect at Oyster Creek.
On May 9, 1987, it was discovered that six temporary variations associated with plant systems did not receive an adequate safety review prior to their installation.
Upon performing another safety
- review, these temporary i
variations were determined to be unacceptable and were revoked.
Four temporary variations were associated with the offgas system and two temporary variations were associated with local level indications for the shell side of the isolation condensers.
Innediate corrective action consisted of isolating from the system all six temporary variations.
Four of the six temporary variations' have subsequently been removed, and the other two will remain isolated pending further evaluation.
The cause of this occurrence was inadequate safety reviews of the temporary variation by. the originator and subsequent reviewers.
The cause and corrective actions associated with the inadequate safety reviews has.been reported in Licensee Event Report 87-021.
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MONELY OPERATING REIORT - JUNE 1987 At the beginning of the report period, Oyster Creek was operating at approximately 640 FMe, limited by increased intake temperatures.
On June 5, due to increased vibration on a condensate pump motor, power was reduced to approximately 385 hMe.
Two condensate pump motors were subsequently balanced, and reactor power was slowly increased beginning June 6.
On June 8, at 608 FMe, increased vibration was again observed.
Over the next few days, power was maintained between 300 and 400 FMe to conduct an investigation and develop a repair method.
On June 11, plant load was further reduced to 220 FMe to perform an MSIV full closure test.
Following resolution of the condensate pump motor vibration, plant load was subsequently increased.
On June 19, power was reduced to 600 FMe to remove damaged ventilation duct turning valves from the Feed Pump Room ventilation system.
Power was subsequently returned to maximum load.
On June 24, due to celgrass accumulation in the intake structure restricting water flow, power was reduced to approximately 305 FMe to stabilize condenser
- vacuum, service water pressure and circulating water pump current.
In addition, a high pressure screen wash pump discharge flange failed, which reduced the ability to remove debris from the intake screens.
On June 25, the eelgrass subsided, and power was increased to 412 bMe.
Following repairs to the high pressure screen wash system, reactor power was increased.
Maximum generator load was achieved on June 26.
Maximum generator load was maintained for the balance of the report period except for brief power reductions to perform turbine valve testing, repair a drain valve, and condenser backwashing.
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OPERATING DATA REPORT OPERATING STATUS 1.
DOCKET:
50-219 (i
2.
REPORTING PERIOD:
JUNE, 1987 3.
ITTILITY CONTACT:
J0llN F1. SEDAR, JR.
609-971-4698 1
4.
LICENSED TlIFRMAL POWER (FMt):
1930 5.
NAMEPLATE RATING (GROSS FMe):
687.5 X 0.8 = 550 6.
DESIGN ELECTRICAL RATING (NET bMe):
650 7.j MAXIMU)! DEPDIDABLE CAPACITY (GROSS FMe):
650 8.
MAXIMUM DEPENDABLE CAPACITY (NET FMe):
620 9.
IF CilANGES OCCUR ABOVE SINCE LAST REPORT, GIVE REASONS:
NONE
- 10. POWER LEVEL TO 1011 01 RESTRICTED, IF ANY (NET bMe):
N/A ll.
REASON FOR RESTRICTION, IF ANY:
NONE MONTil YEAR CUMULATIVE
- 12. REPORT PERIOD IRS 720.0 4343.0 153576.0
- 13. Il00RS RX CRITICAL 720.0 3084.8 97921.3
- 14. RX RESEkVE SilTDWN IRS 00 0.0 918.2
- 15. IIRS GENERATOR ON-LINE 720.0 2961.3 95331.3
- 16. (Jr RESERVE SilTDWN 1RS 0.0 0.0 1208.6
- 17. GROSS TilERM ENER (5601) 1294000 5189304 158145689
- 18. GROSS ELEC ENER (Mini) 427650 1747590 53415835
- 19. NET ELEC ENER (bS01) 411518 1673110 51283187
- 20. Lif SERVICE FACTOR 100.0 68.2 62.1 21.
(Jr AVAIL FACTOR 100.0 68.2 62.9
- 23. (Jr CAP FACTOR (DER NET) 87.9 59.3 51.4
- 24. UT FORCED OlfrAGE RATE 0.0 31.8 11,2 25.
FORCED OUTAGE IRS 0.0 1381.7 12033.5
- 26. SHUTDOWNS SQlEDULED OVER NEXT 6 MONTHS (TYPE, DATE, DURATION):
N/A 27.
IF CURRENTLY SHlfrDOWN ESTIMATED STARTUP TIME:
N/A 1965B
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AVERAGE DAILY P0lfER LEVEL
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NET Mie DOCKET #........
50-219 J
UNIT......... 0Y5fB1 CREEK #1 i
REPORT DATE
...... JULY 1, 1987 CGtPILED BY,......J0llN H. SEDAR, JR.
TELEPfl0NE #......
609-971-4698 l
h0 NTH JUNE, 1987 DAY Mi DAY Mi 1.
614 16, 606 2.
612 17.
609 j
3.
618 18.
617 4.
623 19.
604 5.
618 20.
606 6.
390 21.
604 7.
528 22, 616 8.
589 23.
610 9.
346 24.
611 10, 373 25.
422 11.
465 26.
595 12, 607 27.
595 13.
613 28, 606 14, 612 29.
616 15, 611 30, 610 1968B
Oyster Creek Station #1 Docket No. 50-219 REFUELING INFORMATION - June 1987 Name of Facility:
Oyster Creek Station #1 Scheduled date for next refueling shutdown:
September 1988 l
Scheduled date for restart following refueling: December 1988 Will refueling or resumption of operation thereafter require a Technical Specification change or other license amendment?
No Scheduled date(s) for submi tting proposed licensing action and supporting information:
Important licensing considerations associated with re fueling e.g.,
new or different fuel design or supplier, unreviewed design or performance analysis methods, significant changes in fuel design, new operating procedures:
fuel design and performance 1.
General Electric Fuel Assemblies 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:
2,600 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 (10) racks have been installed to date.
When reracking is completed, discharge capacity tc the spent fuel pool will be available until 1990 refueling outage.
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GPU Nuclear Corporation l
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Post Othee Box 388 l
Route 9 South
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Forked River, New Jersey 08731-0388 l
609 971 4000 j
Writers Direct Dial Number:
I Director Office of Management Information July 15, 1987
]
U.S. Nuclear Regulatory Commission Washington, DC 20555
Dear Sir:
Subject:
Oyster Creek Nuclear Generating Station Docket No. 50-219 Monthly Operating Report l
l 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, wu P
ie er 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 King of Prussia, PA 19406 Mr. Alexander W. Dromerick, Project Manager U.S. Nuclear Regulatory Comminion Division of Reactor Projects I/II 7920 Norfolk Avenue, Phillips Bldg.
$y Bethesda, MD 20014
/
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NRC Resident Inspector g
Oyster Creek Nuclear Generating Station GPU Nuclear Corporahon is a subsidiary of the General Public Utihties Corporahon Q.