ML20237K644
| ML20237K644 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 04/30/1987 |
| From: | Baran R, Fiedler P, Sedar J GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC |
| References | |
| 2060G, NUDOCS 8708190350 | |
| Download: ML20237K644 (11) | |
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s-MONTHLY OPERATING REPORT - APRIL 1987 At the beginning of the report period, Oyster Creek was operating at 656 MWe.
On April 1,
and again on April 7,
brief power reductions were performed to support turbine valve testing.
On April 11, power was reduced to approximately 350 MWe to perform a control rod sequence exchange.
Power was subsequently increased and maintained at approximately 92t to establish xenon equilibrium.
j l
On April 14, as power was being increased, a half-scram signal was received on APRM Channel 4 prior to reaching full power.
The half-scram l
was reset.
APRM Channel 4 was bypassed to prevent spurious half-scram signals and power was held at approximately 92.74.
Investigation revealed APRM Channel 4 had properly responded to a high flux peak low in tne core.
Tuc high flux peak was due to the new rod pattern and was within fuel limits.
On April 16, power was reduced to approximately 80%
i to accommodate a minor rod position adjustment to reduce flux peaks and j
l allow full power op3 ration. Maximum plant load was reached on April 17.
On April 21, power was briefly reduced to 600 MWe to perform turbine valve testing.
On April 22, an acoustic monitor associated with an electromatic relief j
valve (EMRV) was declared inoperable.
Troubleshooting could not identify the cause of failure.
In accordance with Technical Specifications, reactor shutdown commenced on April 23.
All control rods were fully 1
inserted by April 24.
Tne acoustic monitor and its associated line l
driver were subsequently replaced.
Due to a concern with elevated drywell temperatures experienced during recent plant operation, the decision was made to replace the cooling coils on four (4) drywell recirculation fans.
Tne reactor remained shutdown for the balance of the report period.
2060g
MONTHLY OPERATING REPORT-MARCH 1987 The following Licensee Event Reports were submitted during the month of March 1987:
i 1
l Licensee Event Report 50-219/87-007 " Backup Sample Analysis Invalid Due to l
Personnel Error":
i I
On December 31, 1986 a release of processed water was made to the j
envi ronment with the radwaste liquid monitor inoperabl e.
It was i
discovered on January 30, 1987 that the Technical Specification limiting conditions for discharging overboard were not sa ti s fied.
Technical
{
Specifications require that two samples be collected and analyzed, one l
before starting the release and one near the end of the release when radwaste liquid radiation monitors are inoperable.
Both samples were correctly collected, however, the second sample was incorrectly analyzed by a chemi stry technician who inadvertently used the wrong computer program, therefore, the Technical Specification governing the release was J
l nots met.
Changes have been made to the computer program to preclude l
l l
inadvertent selection errors in the future.
\\
Licensee Event Report 50-219/87-008 " Limiting Safety System Setpoint Found in Excess of Tech Spec Limit":
On February 2,1987 at 0347 hours0.00402 days <br />0.0964 hours <br />5.737434e-4 weeks <br />1.320335e-4 months <br /> during surveillance testing the reactor recirculation flow converter scram setpoint was found to be above the Limiting Safety System setpoint listed in the Technical Specifications.
The setpoint is requi red to be less than or equal to 11 7% of rated recirculation flow, and the instrument's setpoint was found to be 118%.
At the time, the reactor was in the RUN mode, producing approximately 96%
power.
The apparent cause of the condition is instrument drif t.
There are two reactor recirculation flow converters, and signals from both are required to scram the reactor, therefore, a reactor scram on high recirculation flow would not have occurred until 118%
rated fl ow.
Innediate corrective action was taken to adjust the setpoint within i
Technical Specification requirements.
A change to the flow converter i
setpoint will be investigated to prevent recurrence.
Licensee Event Report 50-219/87-010 " Electrical Transient Causes Cont o "nent Isolation and Standby Gas Treatment Initiation Due to Design ConfiguratiC." -
On February 23, 1987 a
primary containment i sol ation, secondary containment isolation and Standby Gas Treatment System (SBGTS) initiation occJrred during an automatic electrical bus transfer.
At the time, the reactor was in the SHUTDOWN mode.
The event occurred when heavy snow fall caused a transient in the 34.5 kV distribution lines outside the plant.
The voltage transient caused Vital AC Power Panels to transfer to their al ternate power supply.
The transfer caused reactor protection system (RPS) relays to deenergize causing the containment isolations and SBGTS initiation.
The i solation signal was reset manually.
The safety significance of the event is considered minimal.
Because events of this
Licensee Event Reports l
March 1987 Page 2 nature occurred on June 11, July 29, and July 30 of 1986, an evaluation was performed which revealed that automatic Dus transfer times exceeded relay drop out times for RPS relays.
Several modi fications to correct this condition are being considered.
A follow-up report to LER 86-17, which involved a
similar occurrence, will be subnitted when the l
evaluations are complete.
Licensee Event Report 50-219/87-011 "High RPV Level Turbine Trip / Scram Caused by lost Feedwater Flow Signal Due to Procedural Inadequacy":
On February 14, 1987 at approximately-1301 hours an anticipatory reactor scram occurred as a result of a high reactor water level turbine trip.
The high reactor water level was caused when a technician moved a control room panel wire harness which inadvertently disconnected the "A" feedwater flow signal wire.
A loss of "A" string feedwater flow signal re sul ted.
The Feedwater Level Control System responded to the apparent loss of "A"
feedwater flow by automatically opening the Feedwater Regulating Valves (FRV).
This raised reactor water level until the turbine tripped.
The pressure transient following the turbine stop valve closure, which results in a reactor scram, led to an automatic reactor recirculation pump trip, isolation condenser initiation, and opening of the electromatic relief valves.
Reactor pressure was subsequently controlled using the turbine bypass valves.
The root cause of this event was insufficient procedu al controls over wire termination practices.
The "A" feedwater flow signal wire was the smaller of two different gauge wires in a clamp down type terminal.
This configuration led to the smaller wire becoming loose.
Immediate corrective action was taken to reinstall the feedwater flow r
signal wire.
Terminations in the control room, and others vital to safety and operation were inspected and tightened where necessary.
Procedures will be revised to further control future termination practices.
1 Licensee Event Resort 50-219/87-012 " Inoperable Offgas Drain Line Isolation Valve Caused By De)ris Accumulation Due to Inadequate )reventive Maintenance":
On February 21,
1987 during maintenance activities plant personnel discovered that the air ejector offgas drain line isolation valve was inoperable.
At the time the plant was shut down with reactor coolant temperature less than 212*F.
The valve had failed in the open direction, making it incapable of isolating on an offgas system isolation signal.
The valve was removed and replaced with a similar valve.
The apparent uuse of the valve failure is a buildup of debris over time in the bonnet chamber impeding stem travel.
The safety significance of this condition is minimal because plant procedures require a reactor shutdown on a high radiation signal in the offgas
- line, regardless of offgas system isolation.
Additionally, the main steam isolation valvcs could be closed to rapidily terminate any release from this offgas hold up line.
Future corrective action entails physical examination of the valve on a refueling outage interval and an investigation as to the source and extent of the debris accumulation in the air ejector offgas drain line.
I dmd:(0841 A)
1 MONTHLY OPERATING REPORT APRll 1987 l
The following Licensee Event Reports were submitted during the month of April 1987:
Licensee Event Report 50-219/87-009 " Operation of the Plant with Flow Biased l
Sc ram and Rod Block Setpoints Outside of the Analyzed Region Due to l
Recirculation Loop Backflow":
I On February 7,1987 a loss of generator field to the B recirculation pump (RCP) motor generator (MG) occurred.
The breaker supplying power to the MG failed to trip and no alarm was received to indicate that B RCP was not I
operating.
Reverse flow in the B RCL was incorrectly added to the total recirculation flow signal by the flow summing network.
This caused the flow biased average power range monitor setpoints for the scram and rod block setpoints to be less conservative than allowed by the Technical l
Specifications.
Recognizing that flow through the B recirculation loop
)
was in the reverse direction, the operators closed the B recirculation loop discharge val ve.
The root cause of this event was inadequate preventive maintenance and testing of equipment associated with the recirculation pumps.
Subsequent tra'nsient analyses determined the increased setpoints did not cause the safety limits of the reactor to be violated for any postulated transients or accidents.
The components that I
caused the loss of the generator field were replaced.
The preventive
)
maintenance program for equipment associated with RCPs will be improved.
)
The safety analyses performed for thi s event will be reviewed for i
inclusion in the next annual revision to the Updated Final _ Safety Analysis Report.
1 I
1 Licensee Event Report 50-219/87-013 "SGTS Initiation Caused by Improperly Installed Wire Connector Due to Personnel Error" On March 16, 1987 with the reactor at approximately 100% of rated power, a secondary containment isolation and Standby Gas Treatment System (SGTS) initiation occurred when an electrical connector fell off a recently replaced instrument power supply.
The power supply, which feeds several Area Radiation Monitors, was withdrawn from its panel for a surveillance.
When the power supply was inserted to its normal location the wire connector attached to the rear of the unit fell off causing the event.
The apparent cause of this event is attributed to personnel error in that the instrument technician who installed the new power supplies did not recognize a problem with the exi sting wire pl ug and its new socket connector on the power supply.
Contributing to the cause is the fact that the new supplies were purchased as one for one repl acements yet differences in the socket connectors existed.
No mention of the connector difference was made in the vendor documentation.
The safety significance of this event is considered minimal because the affected safety systems were available and operated as designed for the conditions that existed.
The connectors on this and similar new power supplies have been secured to I
prevent similar problems.
Licensee Event Reports l
April 1987 j
Page 2 l
1 Licensee Event Report 50-219/87-014 "Drywell, Isolation Caused by a Lif ted Lead" On March 3,
- 1987, with the reactor shutdown, an electrical lead termination repai r.
was in progress.
An electrician lifted a lead l
supplying electrical power to the Neutron Monitoring System associated with Reactor Protection System (RPS) channel two (2) and the Containment i
High Range Radiation Monitoring (CHRM) system channel two (2).
This l
caused a half scram and a containment vent and purge valve isolation.
The cause of the event has been attributed to the drawings utilized during the preparation and review )* the detailed instructions which did not reflect recent plant modifications.
The safety significance of this event is considered minimal as the vent and purge valve isolations had no adverse effect on plant systems.
Licensee Event Report 50-219/87-015 "Inoperabl e Intermediate Range Monitors Due to Broken Flexible Connection Caused by Improper Maintenance" On February 18, 1987, at 2335 hours0.027 days <br />0.649 hours <br />0.00386 weeks <br />8.884675e-4 months <br />, while observing source range monitor (SRM) to intermediate range monitor (IRM) overlap during reactor startup, l
two IRMs in one reactor protection trip system (RPS) and one IRM in the other RPS were discovered to be inoperable.
Testing was performed and at 0345 hours0.00399 days <br />0.0958 hours <br />5.704365e-4 weeks <br />1.312725e-4 months <br /> on February 19, 1987, a reactor shutdown was commenced.
At 0516 hours0.00597 days <br />0.143 hours <br />8.531746e-4 weeks <br />1.96338e-4 months <br /> the reactor shutdown was complete.
Subsequent testing revealed that the IRM detector failures were due to a break in the nickel ribbon flexible conductor attached to the detector anode.
The cause of the ribbon breaks was attributed to excessive axial vibration caused by the detector drive mechanisms.
The excessive vibration occurred due to the failure of drive mechanism components as a result of improper assembly and adjustments.
The root cause of the detector failure was improper maintenance.
Seven IRMs and three SRMs were replaced and repairs and tests were performed on the drive mechanisms.
Maintenance procedures will be revised and training will be given to technicians working on IRMs and SRMs.
The preventive maintenance for drive mechanisms will be improved.
Licensee Event Report 50-219/87-016 "Setpoints for Three of Eight Isolation Condenser Pipe Break Sensors Out of Specification De? to Instrument Drif t":
On March 4,
1987 an instrument technician performed the quarterly scheduled surveillance test of the isolation condenser pipe break sensors.
Test resul ts revealed 3 of the 8 sensors had trip setpoints which had drifted above the maximum allowable setpoint specified in the Technical Specifications.
The root cause of the setpoint drift has not been specifically determined, however, there are several possible causes including worn switch linkages, misalignment of switch linkages, and the effects of ambient temperature changes.
An examination of the previous surveillance test results for the isolation condenser pipe break sensors reveals a trend which suggests the sensor's setpoint drifts with ambient temperature changes.
GPUN is evaluating a replacement for the pipe break sensors which would not be susceptible to a setpoint drift problem.
Licensee Event Reports April 1987 Page 3 i
Licensee Event Report 50-219/87-017 " Technical Specification Violation Caused by Inap 3ropriate Removal of Snubbers From Surveillance Program Due to Personne1 Error" On March 11, 1987 plant personnel discovered that two mechanical snubbers in Core Spray System 2 had not been inspected since their installation in 1980, in violation of plant technical specifications.
The snubbers were installed in 1980 to replace two other snubbers in a high radiation area, but their new location was not reflected on piping isometric diagrams.
In November 1983, contractor personnel could not locate the two snubbers as l
listed in the procedure, assumed they did not exist, and they were deleted from the procedure.
The condition was discovered during investigation of l
maintenance work to be performed in the area of the snubbers.
The safety l
significance of the condition is considered minimal because the results of a
prelimina ry piping analysis considering the snubbers to be non-functional indicated that piping stress l evel s did not exceed operability limits.
Innediate corrective actions were taken to inspect the snubbers, include the snubbers in th'e appropriate inspection procedure and initiate a drawing change to correct the piping isometric diagram.
l Licensee Event Report 50-219/87-018
" Reactor Building Ventilation Valve i
Inoperable for Maintenance and Not Secured Closed Due to Personnel Error" On March 20, 1987 an equipment operator isolated the air supply to reactor building ventilation system valve V-28-38.
The valve was being removed from service for maintenance.
The operator turned the handwheel to the closed position and placed an out of service tag on the valve, however, while closing the val ve, he failed to engage the linkage from the handwheel to the valve operator with the engagement pin.
Al though the valve indicated closed, it remained in the open position.
The reactor was in the RUN mode at approximately 99% power at the time.
The condition was discovered on March 25, 1987 when another operator was removing the tags from the valve and restoring it to operable status.
The cause of the event has been attributed to operator error caused by inadequate training.
The safety significance of the event is minimal because the redundant isolation valve was open but operable during the time the first valve was disabled.
Immediate corrective action was taken to ensure both valves were operabl e.
Future corrective actions include operator
- training, providing procedural guidance for manual valve operation, required reading, and providing pins locally for manual valve operation.
dmd:0841A l
l
Oyster Creek Station 01 Docket No. 50-219 REFUELING INFORMATION - April 1987 Name of Facility:
Oyster Creek Station #1 Scheduled date for next refueling shutdown:
N/A l
l Scheduled date for restart following refueling:
Will refueling or resumption of operation thereafter requi re 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 refueli ng 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.
i 2.
Exxon Fuel Assemblies - no major changes have been made nor are there l
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 I
=
l 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:
2,600 l
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 to the spent fuel pool will be available until 1990 refueling outage.
4 l
1
_____-__-_O
l OPERATING DATA REPORT g
OPERATING STATUS g )/ITR 1.
DOCKET:
50-219 2.
REPORTING PERIOD:
APRIL, 1987 3.
UTILITY COtlTACT:
JOHN SEDAR JR.
609-971-4698 4.
LICENSED THERMAL POWER (Kdt):
1930 5.
NAMEPLATE RATING (GROSS MWe):
687.5 X 0.8 = 550 6.
DESIGN ELECTRICAL RATING (NET Kde):
650 7.
MAXIMUM DEPENDABLE CAPACITY (GROSS Kde):
650 8.
MAXIMUM DEPENDABLE CAPACITY (NET Kde):
620 9.
IF CHANGES OCCUR ABOVE SINCE LAST REPORT, GIVE REASONS:
NONE 10.
POWER LEVEL TO WHICH RESTRICTED, IF AtW (NET MWe):
N/A l
11.
REASON FOR RESTRICTION, IF ANY:
NONE l
MONTH YEAR CUMULATIVE 12.
REPORT PERIOD HRS 719.0 2879.0 152112.0 l
13.
HOURS RX CRITICAL 555.0 1958.1 96794.6 14.
RX RESERVE SHTDWN HRS 0.0 0.0 918.2 15.
HRS GENERATOR ON-LINE 554.1 1863.8 94233.8 16.
UT RESERVE SHTDWN HRS 0.0 0.0 1208.6
- 17. GROSS THERM ENER (MWH) 1028000 3215604 156171989
- 18. GROSS ELEC ENER (MWH) 350600 1093200 52761445 19.
NET ELEC ENER (MWH) 337053 1045436 50655513 20.
UT SERVICE FACIOR 77.1 64.7 62.0 21.
IJf AVAIL FACIOR 77.1 64.7 62.7 22.
UT CAP FACIOR (MDC NET) 75.6 58.6 53.7 1
23.
UT CAP FACIOR (DER NET) 72.1 55.9 51.2 l
24.
UT FORCED OUTAGE RATE 22.9 35.3 11.0 25.
FORCED OUTAGE HRS 164.9 1015.2 11667.0 26.
SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, DURATION):
N/A 27.
IF CURRENTLY SHUTDOWN ESTIMATED STARTUP TIME: May 14,19 87 1965B t
I
1 1
l AVERAGE DAILY POWER LEVEL l
NET MWe l
1 7
DOCKET f........
50-219 l
l UNIT........... OYSTER CREEK #1 I
REPORT DATE.
..MAY 5, 1987 j
COMPILED BY....... JOHN SEDAR JR.
TELEPHONE #......
609-971-4698 I
l MONTH APRIL, 1987 l
i 1.
624 16.
584 l
1 2.
637 17, 596 I
l 3.
636 18.
636 I
i 4.
636 19.
636 I
5.
610 20.
636 j
6.
638 21.
631 7.
627 22, 635 1
8.
638 23.
592 9.
638 24.
18 10.
629 25.
0 11.
485 26.
0 12.
573 27.
0 13.
570 28.
0 14, 586 29.
0 15.
587 30.
0 1968B
1
{.
l De S l
GPU Nuclear Corporation l
l 9 Nuclear
- g;<gra88 i
Forked River, New Jersey 08731-0388 1
609 971-4000 I
Writer's Direct Dial Number:
Director May 15, 1987 Office of Management Information U.S. Nuclear Regulatory Commission Washington, DC 20555
Dear Sir:
j
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, 0 ' ster Creek Licensing Manager at (609)971-4643.
r truly o
)
(Peuer L). Fiedler
/
y5 i
Vi;e P esident and Director 0 st b Creek i
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 i
Mr. Alexander W. Dromerick, Project Manager U.S. Nuclear Regulatory Commiss'on Division of Reactor Projects I/II 7920 Norfolk Avenue, Phillips Bldg.
h Bethesda, MD 20014
).
NRC Resident Inspector
[
Oyster Creek Nuclear Generating Station GPU Nuclear Corporation is a subsidiary of tt 3eneral Public Utilities Corporation l
E____________________
.__________m