ML20237K453
| ML20237K453 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 02/28/1987 |
| From: | Baran R, Fiedler P, Notigan D GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC |
| References | |
| NUDOCS 8708190261 | |
| Download: ML20237K453 (8) | |
Text
___
AVERAGE DAILY POWER LEVEL NET Mde
.l DOCKET #........
50-219 UNIT........... OYSTER CREEK #1 REPORT DATE....... MARCH 4, 1987 COMPILED BY....... DONALD V. NOTIGAN l
TELEPHONE #......
609-971-4695 l
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2.
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3.
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563 22.
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624 24.
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10.
488 25.
O 11.
517 26.
0 12.
625 27.
0 13.
630 28.
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MONTilLY OPERATING REPORT - FEBRUARY 1987 l
At the beginning of the report period, Oyster Creek was operating at approximately 652 MWe, limited by condensate discharge header pressure.
On February 7, a rapid load reduction was experienced due to the loss of
'B' reactor recirculation pump.
Plant load was stabilized at j
approximately 550 MWe, The malfunction appeared to be an excitation-problem with 'B' Motor Generator Set.
Troubleshooting is in progress to identify the specific cause of failure.
On February 8, at a plant load of 602 MWe, a reactor shutdown commenced i
after a main steam line radiation monitor was discovered out-of-specifica-
' tion during surveillance testing.,
The monitor was recalibrates and the shutdown was terminated.
Plant load reached a maximum of 655 MWe by February 9.
On February 10, a gaseous radioactive release of unknown origin was discovered in the Turbine Building and plant load was reduced to approximately 485 MWe. An " Unusual Event" was declared at 0610 hours0.00706 days <br />0.169 hours <br />0.00101 weeks <br />2.32105e-4 months <br /> and all notifications were made as reouired.
The gaseous radioactive release was subsequently traced to the Off-Gas Sample System where a leak was located in the line where manual off-gas samples are obtained.
This section of line was isolated and the balance of the Off-Gas Sample System was returned-to-service.
The " Unusual Event" was terminated at-2026 hours.
Plant load was further reduced to 400 MWe to facilitate repairs to a Turbine Building Closed Cooling Water (TBCCW) leak on.'A' condensate pump motor.
On February 11, 'A' condensate pump was returned-to-service and plant load increases commenced.
A plant load of 657 MWe was reached on February 13.
On February 14, while performing a verification of lifted leads (temporary variations) behind control room panels, a full reactor scram occurred following a turbine trip and closure of the turbine stop valves.
The turbine trip resulted from a feedwater transient which was initiated by a wire pulled loose in the feedwater control circuit.
Following the scram, the reacter was brought to a co.1d shutdown condition to accommodate a drywell entry for repairs to an isolation valve in the Hydrogen /0xygen Monitoring System and to repair a faulty ' acoustic monitor associated with a reactor safety valve.
Reactor startup commenced on February 18.
Shortly after achieving criticality, it was noted that there was no detector response on IRM channels 12, 13 and 17.
IRM System I was declared inoperable and in accordance with Technical Specifications, a half-scram was inserted while the problem was investigated.
It was determined that the IRM's could not be readily repaired.
The half-scram was reset and the reactor was then placed in a cold shutdown condition.
All control rods were fully inserted on February 19.
The reactor remained shutdown for the balance of the report period.
4 1505g
MONTHLY OPERATING REPORT-FEBRUARY 1987 i
~
~.
The following Licensee Event Reports were submitted during the month of l
February 1987:
Licensee Event Report 50-219/87-001 " Absence of Neutron Flux Control Rod Block Clamping Ci rcuit Oue to Inconsi stency Between Technical Specifications and Plant Hardware":
On January 2,
1987, while the plant was in the REFUEL mode, it was discovered that there is no clamp ci rcuit to limit the neutron fl ux control rod block setpoint (EIIS System Code JC) to the 108% of rated neutron flux for reactor reci rculation flows of 100% or greater.
The j
condition has been present since initial plant operation and is the result j
of an inconsistency between technical specifications and plant hardware, j
In addition, the existing surveillance procedure required testing the J
control rod block setpoint at 50% and 100% of rated recirculation flow (the normal operating range) but it did not require testing for a control rod block clamp circuit for recirculation flow greater than 100%.
The i
movement of control rods is being controlled admini strative1y when reci rculation flew is at 100% of rated.
A review of all surveillance which fulfill technical specification requirements will be conducted to determine if any similar condition exists.
I Licensee Event Report 50-219/87-002 " Main Steam Isolation Valve Closure Caused by Operator Error":
On January 6,1987 at 0410 hours0.00475 days <br />0.114 hours <br />6.779101e-4 weeks <br />1.56005e-4 months <br />, an inadvertent automatic main steam line j
isolation occurred during startup activities.
Before the event, the j
reactor was critical in the STARTUP mode.
A control room operator had increased power above the pnint of adding heat and had established a heat up rate.
When a control rod was withdrawn to maintain the heat up rate, the indicated power on the adjacent IRM increased rapidly.
The control room operator responded to the increased power level by ranging that IRM into range 10.
The Reactor Protection System responded by shutting the MSIVs and all other valves required to shut on a reactor isolation since i
reactor pressure was less than or equal to 825 psig.
The cause of the a
event was determined to be operator error in not realizing that one IRM was in range 9 with reactor pressure less than or equal to 825 psig and W
then ranging to range 10.
Corrective action to be taken in the future j
includes operator training on this event.
j l
Licensee Event Report 50-219/87-003 " Standby Gas Treatment System Initiation Caused by Power Supply Perturbation":
1 On January 7,1987 the Reactor Building Ventilation System isolated and j
the Standby Gas Treatment System initiated when a
power supply perturbation tripped the reactor building ventilation radiation monitors.
I At the time of the event the plant was running at approximately 32% rated j
power with reactor pressure at 1001 psig and temperature at 526*F.
The investigation revealed some Area Radiation Monitor (ARM) ribbon cables were nicked, exposing the cables to possible shorting problems.
The power supplies in the ARM trip circuits have also shown some degradation from i
age and use in the past.
The degraded equipment condition due to aging is i
considered the probable root cause of the event.
Corrective action taken l
was to repair the ARM ribbon cables found nicked.
_-__-j
~ '
.. Licensee Event Reports February 1987 s
Page__2 t
Licensee Event Report 50-219/87-004 " Technical Specification Violation Caused by Improper Removal of Equipment From Service Due to Personnel Error":
On January 12, 1987 emergency service' water (ESW) pump D was-tagged out of service for corrective maintenance while emergency diesel generator'#1.was inoperable for surveillance.
This occurrence ' was in violation. of ' the Oyster Creek Technical Specifications in that Technical Specifications-require that when one EDG is inoperable, all engineered safety features fed by the operable EDG' must. be ' operable.-
The root cause of the occurrence was personnel ' error in that operations management, the senior!
reactor operator, and 'the shift technical advisor on shift did not recognize the Technical Specification' violation in the planned tag-out of equipment.
Corrective action consisted of returning ESW pump D to service.
This. event will be discussed in iicensed operator training and' will be made required reading for operation 1 management personnel.
Licensee Event Report 50-219/87-005 "High Flux Scraa During Recirculation Pump Start Due to Discharge Valve Partially Open":
On January 16, 1987, a high flux reactor scram _ occurred due to increased recirculation flow.
At the time, "E"
recirculation pump-was undergoing testing which required starting the pump.
The pump discharge valve had been
- closed, however, as ~ the pump was
The cause 'of the event is attributed to a partially open discharge valve.. The root cause of the -
event was a low torque switch setting which controls the closing thrust of the discharge valve.
The closing thrust of the "E"
recirculation pump discharge valve was measured and the torque switch setting was raised to its proper value.
Torque switch settings for the remaining four' discharge valves will be similarly reset.
Licensee Event Report 50-219/87-006 " Technical Specification Violation' Caused by Improper Storage of Higher Enrichment Fuel Due to Personnel Error":
Oyster Creek Technical Specifications 5.3.1(C) specify that the fuel stored in the fuel pool storage racks shall not exceed a maximum average planar enrichment of 3,01 wt% U-235 Contrary to the above, reload fuel bundles supplied by General Electric Company (GE) having an average planar enrichment of 3.19% U-235 were temporarily stored in the fuel pool during the 11P. outage in 1986. The cause of the event is attributed to' personnel error in not performing a thorough safety analysis for storage of the new fuel and in not recognizing a conflict with the Technical-Specifications prior to fuel storage in the spent fuel pool.
Corrective actions will consi st of revi sing the refueling procedures, revi sing the Te hnical Specifications to raise the enrichment limitations on stored fuel, and reviewing the occurrence with engineering personnel.
dam:(0841A) l l
o
Oyster Creek Station #1 Docket No. 50-219 l
REFUELING INFORMATION - FEBRUARY, 1987 I
l Name of Facility: Oyster Creek Station #1 I
l Scheduled date for next refueling shutdown: N/A l
\\
l' Scheduled date for restart following refueling:
Will refueling or resumption of operation thereafter require a Technical Specification change or other license amendment?
00 Scheduled date(s) for submitting proposed licensing action and supporting information:
Iinportant 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.
z 1
- 2. Exxon Puel Assemblies - no major changes have been made nor are there any ar.ticipated.
l l
)
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:
2600 The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licenscd capacity:
Reracking of the fuel pool is in progress. Six (6) out of ten (10) 1 racks have been installed to date. When reracking is completed, I
discharge capacity to the spent fuel pool will be available until 1990 refueling outage.
1619B
1 OPERATING DATA REPORT
~ o OPERATING STATUS 1.
DOCKET:
50-219 2.
REPORTING PERIOD:
FEBRUARY, 1987 I
3.
UTILITY CONTACT:
DONALD V. NOTIGAN 609-971-4695 4.
LICENSED THERMAL PCrdER (Kdt):
1930 j
i 5.
NAMEPLATE RATING (GROSS Kde):
687.5 X 0.8 = 550 6.
DESIGN ELECTRICAL RATING (NET Kde):
650
]
f 7.
MAXIMUM DEPENDABLE CAPACITY (GROSS Kde):
650 8.
MAXIMUM DEPENDABLE CAPACITY (NET Kde):
620 j
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 672.0 1416.0 150649.0 13.
HOURS RX CRITICAL 325.3 861.6 95698.1 l
- 14. RX RESERVE SHTDWN HRS 0.0 0.0 918.2 15.
HRS GENERATOR ON-LINE 325.0 804.2 93174.2 l
16.
UT RESERVE SHTDWN HRS 0.0 0.0 1208.6
- 17. GROSS THERM ENER (MWH) 580200 1233304 154189689
)
l 18.
GROSS ELEC ENER (MWH) 202670 417530 52085775
- 19. NET ELEC ENER (MWH) 193225 396649 50006726 20.
UT SERVICE FACI'OR 48.4 56.8 61.8
(
21.
UT AVAIL FACIOR 48.4 56.8 62.7 22.
UT CAP FACIOR (MDC NET) 46.4 45.2 53.5
UT FORCED OUTAGE RATE 51.6 43.2 10.8 25.
FORCED OUTAGE HRS 347.0 611.8 11263.6 26.
SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, DURATION):
N/A 27.
IF CURRENTLY SHUTDOWN ESTIMATED STARTUP TIME:
March 9, 1987 1965B A
{L j
y.
GPU Niclear Corporation l
1 U Nuclear
- e n e 388 Forked River, New Jersey 08731-0388 609 971-4000 Writer's Direct Dial Number
March 17, 1987 Director Office of Management Information U.S. Nuclear Regulatory Commission Washington, DC 20555 j
)
Dear Sir-Subj ect: 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,
^
I t
iedier l
Vice President and Director l
i Oyster Creek 1
1 PBF:KB: dam (0841A) i Enclosures I
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 l
Mr. Jack N. Donohew, Jr.
I U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue, Phillips Bldg.
Bethesda, MD 20014 Q
'Y NRC Resident Inspector Oyster Creek Nuclear Generating Station y
g GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation
_ - _