ML20099K006
| ML20099K006 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 01/31/1985 |
| From: | Baran R, Fiedler P GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF RESOURCE MANAGEMENT (ORM) |
| References | |
| NUDOCS 8503200138 | |
| Download: ML20099K006 (10) | |
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MONTHLY OPERATIONS REPORT JANUARY 1985 During the entire report period, Oyster Creek remained in operation and generated a total of 467,650 megawatt hours of electricity.
From January 9, 1985 to the end of the report period an average (thermal) power level of 97.5% to 98% was maintained.
On January 1, 1985, load was increased to 92% thermal power (618 MWe).
At this time, extreme vibration was noted on the main turbine hydraulic control enclosure.
Load was decreased to a generator load of 592 MWe at which time the vibration stopped.
The M&C Department adjusted the relief valve setting on the suction of the booster pump on the lube oil system.
Load was increased slowly, and at approximately 600 MWe, the vibration was again noted.
A support (jacking device) was installed below the hydraulic enclosure and the vibration problem stopped.
Further investigation of this problem is still pending.
Power level was increased over a several day period to 98% thermal power, which was reached on January 8, 1985.
The Augmented Off Gas Facility (A0G) was placed in service on January 4, 1985.
Replacement of the suction strainer on No. 2 RBCCW pump was completed on January 6,1985, and the pump was placed back in service.
A six (6) shift Operations schedule went into effect on January 7, 1985.
Torus to drywell vacuum breaker V-26-5 failed its associated surveillance test on January 8, 1985 when it did not close freely.
The valve was secured closed per technical specifications.
On January 9, 1985, due to increased leakage from V-1-34 and No. I flash tank manway, power was decreased and the second stage steam reheaters were taken out of service to facilitate repairs.
Power was decreased to 89% power for repairs and returned to 98% power after the repairs were completed.
Repairs (sealant injection) were successful on the manway and flash tank.
Some Icakage still exists on the flash tank manway, but it has been reduced substantially.
Additional repairs (sealant injection) had to be made over the report period to keep the leakage at a manageable rate.
The spring clip on Cleanup System pressure control valve ND-ll controller was reinstalled.
Pressure control with the valve has been very good.
A cleanup filter was placed in service on January 9, 1985.
Governor repairs and radiator replacement on No.
1 fire diesel were completed on January 10; however, upon testing, it was noted that the speed (RPM's) of the diesel exceeded the procedure acceptance criteria (high).
The diesel vendor was called in and the speed was adjusted to within proper limits on January 15.
During the time period the speed problem was being resolved, the fire diesel was considered operable per Plant Engineering.
8503200138 850131 DR ADOCK 05000219 PDR
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M:nthly Oparaticus Raport Page 2 l
i During the'. monthly report period, the average drywell unidentified leak I
rate plus the unidentified leak into the torus increased from 2.6 gpm to 4 gpm.. Subsequent to this report period, the major contributor to the unidentified leak rate was determined to be a pressure seal on Feedwater System valve V-2-36.
I Fuel pool heat exchanger "A" remained out of service during the report period.
Repairs included rolling of the tubes and plugging approximately 66% of the tubes (this proved unsuccessful).
Further investigation is presently on hold due to.the present mini-outage which started February 2,1985. A new heat exchanger is also being ordered.
Drywell bulk temperature has been running approximately 1370F.
I Operations Department has taken action to reduce the temperature to below I
1350F, the established limit; however, success at substantial reduction in temperatures has been limited.
This problem is currently being
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addressed and causes evaluated by Tech. Functions and Plant Engineering.
Due to severe cold weather conditions, pressure sensing lines and transmitters on various systems at the Intake Structure froze.
A heat tracing system has been installed and presently is in use.
Extreme cold weather early in the report period caused multiple problems with the Intake screens.
Nos. 2 and 3 intake screens became inoperable.
Portable heaters were installed to thaw out the ice inside the screens.
j No. 3 screen was returned to operation on January 22, 1985.. No. 2 screen is still out of service and will be pulled after the present mini-outage.
Intake Structure thermal dilution gates are being opened to help prevent icing problems.
Three of the gates required maintenance to facilitate operation.
j Repairs to Nos. I and 2 fire system deluge piping were completed after they were damaged by freezing conditions.
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Environmental Qualification repairs were commenced during the report period on Core Spray System valve limitorques (new torque switches).
I The muffler on fire diesel. No. 2 was replaced on January 16, 1985.
A high speed problem was again corrected on No. 2 fire diesel on January 25, 1985.
A high speed logic problem with No.
1 TIP machine was resolved.
additional shielding was also installed around the TIP - Room to reduce radiation levels in the area.
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M:nthly Oparctieno Raport Page 3 A spiking problem with IRM 14 has been identified as being located in the drywell and will be addressed if and when a mini-outage occurs.
On January 21, 1985, a 5% voltage reduction was experienced on the PJM electrical grid due to heavy power demands.
Non-essential loads were j
secured at the plant until the situation was terminated by the System.
j The voltage reduction lasted approximately two hours.
j Temperature indicators (TK) for reactor safety valves NR28A and NR28W were repaired during the report period.
The thermocouples act as backup position indication for the safety valves.
The acoustic monitors for NR28L and NR28G were also repaired and satisfactorily tested.
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No. I reactor protection system (RPS) motor generator set is presently out of service to replace its flywheel bearings.
RPS No. I was supplied i
power from the standby transformer during maintenance (returned to service on February 4, 1985).
Corrective maintenance to plant recorders continued during the report period due to recurring chart take up problems.
l Mechanical repairs to the limitorque operator on RBCCW valve V-5-106 (RBCCW to the shutdown cooling system heat exchanger) were completed on January 30, 1985.
The limitorque will be connected electrically when j
manpower permits (mini outage).
i Leakage from the "B"
feedwater string minimum flow valve has increased during the report period. A flange leak on V-2-11 ("B" feedwater string
]
outlet valve) has also developed (outage items for repair).
Gaskets on the instrument air post filters failed twice during the report period (January 22, 1985 and January 24, 1985).
In both cases, air pressure decreased to around 60 psig before the filters could be bypassed to re-establish proper pressure. After the second failure, a new type of gasket material was used and the seating surfaces on the filters were machined. So far, no additional problems were encountered.
Air compressor No. 2 remained out of service during most of the report period after its after-cooler was replaced. Repair of air leaks, cooling water leaks and resolution of unloading problems delayed placing the unit back in' service until January 27, 1985.
Diesel fire pump No. I was declared inoperable on January 30, 1985 due to a cracked drain valve that was discovered on its associated fuel oil tank. Maintenance was completed in early February.
On January 30, 1985, the Core Engineering group notified the Operations Department that the computer (PSMS) was under predicting the MAPLHGR (Maximum Average Planar Linear Heat Generation Rate) limit.
The purpose i
of monitoring MAPLHGR is to limit the amount of decay heat stored in a fuel assembly which could lead to fuel cladding failure in the event of a loss of coolant accident.
In order to bring this limit-within specification, power was decreased from 96.7% to 88.3% (thermal) until
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M:nthly Opsraticus R port Pegs 4 a new rod pattern was established.
Power was then increased to 96.6%.
The power reduction and subsequent increase took about two (2) hours.
Technical Functions discovered a design deficiency with the core spray booster purap (CSBP) logic.
Pressure sensors (RV-40s), are located at the discharge of the CSBPs.
If sufficient discharge pressure (230 psig) is not established within 5 seconds af ter a CSBP start signal is received, the start signal to the primary CSBP (NZO3A/NZ03B) is automatically terminated and a _ start signal to the backup CSBP (NZO3C/NZO3D) is i
initiated.
The Technical Functions group calculated that during a rapid i
depressurization (A.D.S.
initiation),
CSBP discharge pressure will degrade to the point that the. logic trip will occur.
Due to the problem identified above, a reactor shutdown was coaumenced the evening of February 1, 1985.
New differential pressure switches will be installed across the CSBPs and will be tied into the CSBP start logic in place of j
the RV-40 sensors.
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Monthly Opsrating Report Page 5 r
The following Licensee Event Reports were submitted during the month of January 1985:.
I Licensee Event Report 50-219/84 On 12/3/84, an automatic reactor scram occurred due to low condenser vacuum.
A plant heatup was in progress, reactor pressure was 586 psig and power was 100 MWt.
Establishing proper i
condenser vacuum had been difficult due to mechanical vacuum pump problems, but vacuum had been established and was being increased.
Low condenser vacuum scram alams were tripped, but scram function was bypassed by reactor pressure being less than 600 psig and mode switch in STARTUP.
When reactor pressure reached 586 psig, the low pressure bypass of the low vacuum scram function cleared and a full automatic scram resulted.
Failure of the operators to realize the significance of the low vacuum scram alams and procedural 1
inadequacies contributed to the occurrence of the event.
Startup procedures have been revised and other corrective. actions will be implemented to prevent l
a similar occurrence in the future.
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Licensee Event Report 50-219/84 During performance of main steam isolation valve closure tests and in-service tests, main steam drain valves V-1-106,107,- and 110 failed to operate when given appropriate signals.
The apparent cause of occurrence is due to torque switch opening which interrupted j
operation of the valves.
All valves were _ deactivated and secured in their 1
isolation position as required by the Technical Specifications.
The valves 1
will be tested and rocaired at the next operability or IST testing, which will take place during toe next scheduled shutdown.
Licensee Event Report 50-219/84 During nomal plant operation, while perfoming the main steam line low pressure functional test, the main steam line isolation valve isolation sensor switches RE 238 and RE 230 tripped at i
values lower than specified in Tech Specs.
The switches were reset to trip l
within the desired setpoint limit.
9 Licensee Event Report 50-219/82-48, Rev. 1 Violation of Technical l
Specification 3.1. A when the reactor water level instrumentation for 1 channel in each Reactor Protection System and 1 channel in each of several safety j
systems were rendered inoperable as a result of the loss of reference column head.
The cause of erroneous vessel water level indication was a decrease in referencr: leg head.
The reference leg for the affected instruments was filled 4
j and restored as operable.
A test program was implemented and resulted in j
replacement of a valve manifold and repacking of two (2) root valves.
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Licensee Event Report 50-219/83-22, Rev.1 - Two mechanical snubbers were found to be inoperable during Stroke testing as required by IE Bulletin 81-01.
These snubbers were located on the Main Steam System and the Core i
Spray System.
The two failed snubbers were replaced with new operable i
replacement units.
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50-219 UNITSil0TDOWNS AND POWER REDUCTIONS UNITNAME Oyster Creek DATE 2-8-85 COMPLETED BY R.
Baran REPORT MON'DI January 1985 TELEPHONE 971-4640 "L
E 3 "g
.., A Licensee j's;,
- L Cause & Corrective 3g
_8 s E Event g
o-3 Action to No.
Date a-F yE E
Ji g g Report #
mU
$O Prevent Recurrence 6
Q 31 2-11-83 S
15138. LC 1
N/A ZZ ZZZZZZ Start of the 1983/84 Refueling and Maintenance Outage.
i 32 11/4/84 F
479. 9A 1
84.28 SH VALVEX Two EMRVs failed to operate prr'pe rly.
33 11'/30/84 S
139..3B 2
N/A ZZ ZZZZZZ Manual scram to test the scram discharge volume.
34 12/15/84 F
0 B
4 N/A ZZ ZZZZZZ Power reduction of 20% power to check condensate pump performan< :e.
1 35 12/20/84 F
0 B
4 N/A ZZ ZZZZZZ Power reduction 20% power to
' repair the flash tank manway.
s 1.
2 3
4 F: Forced Reason:
Method:
Exhibit G-Instructions S: Scheduled A-Equipment Failure (Explain) 1-Manual for Preparation of Data B-Maintenance of Test 2-Manual Scram.
Entry Sheets for Licensee C-Refueling 3-Automatic Scram.
Event Report (LER) File (NUREG-D-Regularory Restriction 4-Other (Explain) 0161)
E-Operator Training & License Examina ba F-Administrative G-Operational Error (Exphin) 5 Exhibit 1 - Same Source r
(9/77)
Il-Other (Explain)
- s OPERATING DATA REPORT OPERATING STATUS 1.
. DOCKET:
50-219 2.'
REPORTING PERIOD:
January, 1985 3.
UTILITY CONTACT:
JOSEPH R. MOLNAR 609-971-4699 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 TO WHICH RESTRICTED, IF ANY (NET MWe): N/A 11.
REASON FOR RESTRICTION, IF ANY: NONE MONTH YEAR CUMULATIVE 12.
REPORT PERIOD HRS 744.0 744.0 132457.0 13.
HOURS RX CRITICAL 744.0 744.'
86371.9 14 RX RESERVE SHTDWN HRS 0.0 0.0 469.7 15.
HRS GENERATOR ON-LINE 744.0 744.0 84280.7 16.
UT RESERVE SHTDWN HRS 0.0 0.0 2.7 17.
GROSSTHERMENEkt(MWH) 1383000 1383000 138645329 18.
- GROSS ELEC ENER (MWH) 484970 484970 46867965 19.
' NET ELEC ENER (MWH) 467650 467650 45030215 20.
UT SERVICE FACTOR 100.0
.100.0 63.6 21.
UT AVAIL FACTOR 100.0 100.0 63.6 22.
UT CAP FACTOR (MDC NET) 101.4 101.4 54.8 23..
UT CAP FACTOR (DER NET) 96.7 96.7 52.3 24.
UT FORCED OUTAGE RATE 0.0 0.0 9.6 25.
FORCED OUTAGE HRS 0.0-0.0 8951.1 26.
SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, DURATION): N/A 27.
IF CURRENTLY SHUTDOWN ESTIMATED STARTUP TLME:
2/14/85
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AVERAGd DAILY POWER LEVEL NET MWe DOCKET f........
50-219 UNIT.......... 0yster Creek #1 REPORT DATE....... FEBRUARY 1985 COMPILED BY,...
. THOMAS J. MCKINNEY TELEPHONE #...... 609-971-4456 MONTH JANUARY, 1985 DAY MW DAY MW 1.
589 16.
637 2.
573 17.
634 3.
600 18.
640 l
i 4.
607 19.
639 5.
629 20.
638 6.
632 21.
636 7.
634 22.
639 8.
637 23.
639 9.
617 24, 638 10.
637 25.
639 11.
640 26.
640 12.
634 27.
635 13.
635 28.
630 14.
t.'
29.
629 15.
022
- 30.
626 31.
623 1
)
Oyster Creek Station #1 Dockst No. 50-219 REFUELING INFORMATION - January, 1985 Name of Facility: Oyster Creek Station #1 Scheduled date for next refueling shutdown: November 30, 1985 Scheduled date for restart following refueling: June 1,1986 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:
June, 1985 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 =
994 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. Three out of ten (10) racks have been installed to date.
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GPU Nuc ear Constation NggIgf Post Office Box 388 Route 9 South Forked River New Jersey 08731-0388 609 971 4030 Writer's Direct Dial Number:
February 15, 1985 Director 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. Drew Holland at (609) 971-4643.
Very truly yours, O
sa amW-P ter
'iedler Vice President and Director Oyster Creek PBF:KB: dam 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 NRC Resident Inspector jfhh Oyster Creek Nuclear Generating Station l
Forked River, NJ 08731 e I GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation