ML20207F524
| ML20207F524 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 09/30/1986 |
| From: | Fiedler P, Notigan D GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF RESOURCE MANAGEMENT (ORM) |
| References | |
| NUDOCS 8701060107 | |
| Download: ML20207F524 (8) | |
Text
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OPERATING MTA REPORT OPERATING STATUS 1.
DOCKET:
50-219 2.
REPORTING PERIOD:
SEPTEMBER, 1986 3.
UTILITY CONTACT:
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 DEPENMBLE CAPACITY (NET MWe):
620 9.
IF CHANGES OCCUR ABOVE SINCE LAST REPORT, GIVE REASONS:
NONE 10.
POWER LEVEL TO WHICH RESTRICTED, IF ANY (NET MWe):
N/A 11.
REASON FOR RESTRICTION, IF ANY:
NONE MONTH YEAR CUMULATIVE 12.
REPORT PERIOD HRS 720.0 6551.0 147024.0
- 13.. HOURS RX CRITICAL 0.0 2311.5 94758.9 14.
RX RESERVE SHTDWN HRS 0.0 448.5 918.2 15.
HRS GENERATOR ON-LINE 0.0 2273.9 92333.0 16.
UT RESERVE SHTDWN HRS 0.0 452.8 1208.6 17.
GROSS THERM ENER (MWH) 0 4074338 152951807 18.
GROSS ELEC ENER (MWH) 0 1371480 51662165 19.
NET ELEC ENER (MWH)
-1779 1307537 49616138 I
l 20.
UT SERVICE FACTOR 0.0 34.7 62.8 21.
UT AVAIL FACTOR 0.0 41.6 63.6 22.
UT CAP FACTOR (MDC NET) 0.0 32.2 54.4
UT FORCED OUTAGE RATE 0.0 6.3 10.3 25.
FORCED OUTAGE HRS 0.0 152.3 10587.9 26.
SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, DURATION):
N/A 27.
IF CURRENTLY SHUTDOWN ESTIMATED STARTUP TIME:
Nov ember 21, 1986 g/
8701060107 860930 I
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DOCKETNO.
50-219
-a UNITSilUTDOWNS AND POWER REDUCHONS UNITNAME Oyster Creek 4
DATE Oct.
2, 1986 CO.MPLETED RY R.
Baran REPORT MONTII September 1986 TELEPIIONE 971-4640 4
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4 F: Forced Reason:
Method:
Exhibit G. Instructions 4
S: Scheduled A-Equipment Failure (Explain) 1-Manual for Preparation of Data B-Maintenance or Test 2 Manual Scram.
Entry Sheets for Licensee C.Refu: ling 3-Automatic Scram.
Event Report (LER) File (NUREG-D-Regulatory Restriction 4-Other (Explain) 0161)
E-Op:rator Tiaini..g1. Ucense Ex::mination G. Operational Erm (Explain) 5 F-Administrativ:
Exhibit 1-Same Source (9/U)
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- r AVERAGE MILY POWER LEVEL NET MWe
- DOCKET f........ 50-219 UNIT......... ~. 0YSTER C REEK # 1 REPORT MTE...
. 0CTOBER 3, 1986 COMPILED BY....... DONALD V. NOTIGAN TELEPHONE f...... 609-971-4695 MONTH SEPTEMBER, 1986 MY-MW MY MW 1.
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g GPU Nuclear Corporation Qggf Post Office Box 388 Route 9 South Forked River,New Jersey 08731-0388 609 971-4000 Writer's Direct Dial Number:
Director October 15, 1986 Office of Management Information U.S. Nuclear Regulatory Commission Washington, DC 20555
Dear Sir:
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,
.A P
iedler Vice President and Director l
Oyster Creek l
l PBF:BP: dam (0841A)
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 Inspector Oyster Creek Nuclear Generating Station l
GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation
t
4 MONDILY OPERATING REPORT - SEPTEMBER 1986 Throughout the report period,. Oyster Creek ' remained shutdown for the Cycle 11 Refueling / Maintenance / Modifications Outage.
On September 6, the Fire Suppression (deluge) System was inadvertently actuated in the. Cable Spreading Room (CSR) below the control room. This resulted in wetting down various pieces of-electrical equipment, including panel ER2 which provides control rod position indication.
Due to loss of control rod position indication, refueling. activities were suspended while the condition of affected equipment was analyzed. Due to potential loss of other vital-equipment in the CSR, an " Unusual Event" was declared and required notifications were made.
Circuitry cards in panel ER2 and other equipment were subsequently dried and following the restoration of rod position indication, clearing of rod blocks and functional tests, reactor refueling resumed (September 9). A critione of the ~ event revealed that the Fire Suppression System actuation was caused by a heat gun being used to apply wire marker sleeves in the vicinity of a smoke detector.~
' Reactor refueling was completed on September 14.
Low Power Physics testing (shutdown margin measurement and demonstration) completed on September 24 following completion of surveillance testing and other associated prereauisites.
Results of the tests conducted were consistent with analytical predictions.
Standby Gas Treatment System I was declared inoperable on August 27 after failing the charcoal absorption test.
The charcoal absorber cartridge was subsequently replaced.. As a result of continued failures following cartridge _ replacement, vendor test eouipment -was used in parallel with plant equipment which verified test failures were due to faulty plant test equipment.
Testing was satisfactorily completed on September 18.
Operability testing was completed on September 19 and the system returned to operable status.
On September 19, a full reactor scram signal was experienced as a result of 1A2 bus being inadvertently de-energized while IB2 was being removed from service.
Power from -1A2 bus was immediately restored.
A critioue of the event was conducted by the Plant Operations Department.
1147g
MONTHLY OPERATING REPORT-SEPTEMBER 1986 The following Licensee Event Reports were submitted during the month of September 1986:
Licensee Event Report 50-219/86-018 - Secondary Containment Leak Rate Test failed to meet tne acceptance criteria because a cap on the service water overboard line in the seal-well had fallen off.
The cap was reinstalled and the subsequent leak rate test was successful.. The plant was in a
refueling / maintenance outage, -with the reactor defueled.
The Service Water System was out of service and drained to allow replacement of the Reactor Building Closed Cooling Water (RBCCW) heat exchanger.
The service water
. overboard line was capped with a temporary cap in the seal well, outside the reactor building.
The safety significance of the event is considered minimal in light of the plant conditions at the time of the event.
The equipment control / temporary variation procedure 'will be reviewed to determine if any revisions or clarifications are required.
Licensee Event Report 50-219/86-019 - On August 9, 1986, at 1052 hours0.0122 days <br />0.292 hours <br />0.00174 weeks <br />4.00286e-4 months <br /> while the reactor was in the REFUEL mode with the core off-loaded, the Standby Gas Treatment System (SBGTS) initiated and the drywell sump and torus vent valves isolated when an electrical short occurred inside a Control Room panel.
Control Room operators received indication that both SBSTS trains had automatically initiated, and. that drywell sump and torus vent valves had isolated.
An investigation revealed that one end of an electrical jumper inside a Control Room panel had fallen off, and caused a short to ground.
The short blew two fuses in the Reactor Protection System (RPS), initiating the event.-
The fuses were replaced, SBGTS operation was secured, the valve isolations were reset, and the -jumper was removed at 1150 hours0.0133 days <br />0.319 hours <br />0.0019 weeks <br />4.37575e-4 months <br />.
The cause has been attributed to personnel error, resulting from work inside the Control Room panels.
The safety significance of this event is considered minimal.
l Corrective action will be taken to require temporary variation removal before job closeout by supervisors.
Licensee Event Report 50-219/86-020 - On August 11, 1986 at approximately 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br />, inspections of discs in control rod hydraulic control unit (HCU) manual inlet and outlet isolations valves (Y-101 and V-102) were in progress.
- V-102' for HCU 18-15 was found to 'have one of two ears completely broken from the wedge on the valve disc. This failure could result in the valve not being full open which could impair the exhaust of water from above the control rod piston during a scram or insertion.
This failure could also reduce flow to the top of the piston during rod withdrawal.
The cause of the failure was determined to be intergranular stress corrosion cracking which was described in General Electric Service Information Letter 41 9.
This condition is considered potentially generic.
The root cause of the failure was judged to be material choice as well as disc geometry.
The valve was made from AlS1 type 420 stainless steel heat treated to 500 Brinell.
The safety significance of this event is considered minimal, based on the past scram times of the affected control rod drive.
The immediate corrective actions were to replace l
the valve disc and stem and continue inspections of all V-102 valve discs.
1 l
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MONTHLY OPERATING REPORT - SEPTEMBER 1986
^
Lic:ns2 Event R: port - Page 2 Licensee Event Re) ort 50-219/86-022 - A discrepancy was noted between the installation of 80 out of 13/ Control Rod Drive (CRD) Hydraulic Control Units (HCU) and the installation drawing.
An analysis shows that the units in question no not meet the seismic requirement defined by Regulatory Guide 1.26 or the original design basis of the plant for Seismic Class I equipment.
The HCUs direct high pressure water from the CRD pumps for nonnal CRD movement and act as a high pressure water source for the reactor scram function.
If the units were rendered inoperable by a seismic event the Standby Liquid Control system would be available for reactor shutdown.
Additionally, if the piping attached to the HCUs were ruptured, a small break loss of coolant accident would result inside the Secondary Containment due to reactor water leaking past the CRD seals and out of the break.
The improperly installed HCUs will be modified to meet seismic criteria as defined in Regulatory Guide 1.26.
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W Oyster Creek Station #1 Docket No. 50-219 REFUELING INFORMATION - SEPTEMBER, 1986 Name of Facility: Oyster Creek Station #1 j
Scheduled date for next refueling shutdown: Currently in Cycle 11 Refueling Outage Scheduled date for restart following refueling:
Nov ember 21, 1986 Will refueling or resumption of operation thereaf ter require a Technical Specification change or other license amendment?
Yes Scheduled date(s) for submitting proposed licensing action and supporting information:
Submitted June 17, 1986 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, in number of fuel assemblies:
l 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 cf 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.
1619B