05000278/LER-1997-001, :on 970122,determined Plant Thermal Power Exceeded 100% Due to Out of Tolerance Test Equipment.Retired Test Apparatus from Future Use & Will Review and Revise Procedures for Feedwater Temp Instruments
| ML20138L653 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 02/19/1997 |
| From: | Geoffrey Edwards, Lengyel G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LER-97-001, LER-97-1, NUDOCS 9702240466 | |
| Download: ML20138L653 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 2781997001R00 - NRC Website | |
text
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Garrett D. Edwards Plant Manager i
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F%ach Bottom Atonnc Power Station l
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PECO NUCLEAR nm ee c-eev 1848 lay Road A UNir or PICO EN/RGY De:ta, PA 17314-9032 717 456 4244 February 19, 1997 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 Docket Nos. 50-278
SUBJECT:
Licensee Event Report, Peach Bottom Atomic Power Station Unit 3 This LER concerns a condition prohibited by Technical Specifications due to a slight increase in thermal power.
Reference:
Docket No. 50-278 Report Number:
3-97-01 Revision Number:
00 Event Date:
10/22/95 Discovery Date:
01/21/97 Report Date:
02/19/97 Facility:
Peach Bottom Atomic Power Station i
1848 Lay Road, Delta, PA 17314 This LER is being submitted pursuant to the requirements of 10 CFR 50.73(a)(2)(i)(B).
Sincerely, l
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GDE\\DJF:djf enclosure cc:
W. T. Henrick, Public Service Electric & Gas R. R. Janati, Commonwealth of Pennsylvania INPO Records Center H. J. Miller, US NRC, Administrator, Region i R. I. McLean, State of Maryland W. L. Schmidt, US NRC, Senior Resident Inspector 9
A. F. Kirby Ill, DelMarVa Power
'40037 J. A. Isabella, VP - Atlantic Electric
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Qg 9702240466 970219
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On 1/21/97, at 1450 hours0.0168 days <br />0.403 hours <br />0.0024 weeks <br />5.51725e-4 months <br />, it was determined that the plant thermal power level exceeded 100 percent by approximately 0.6 percent due to four feedwater temperature instruments being non-conservatively recalibrated on 6/13/95. These instruments provide input into the overall core thermal power calculation. Once identified, operating personnel promptly reduced reactor power. The non-conservative calibration of the instruments was due to the use of out-of-tolerance test equipment. When the out-of-tolerance test equipment was reported to station personnel in July 1996, no recalibrations were performed. This was due to a lack of specific information available to the reviewer concerning the function of the feedwater temperature instruments. Subsequent investigation has revealed that the test equipment had a history of internal component failures although the failures were on different components within the test equipment. A contributing cause involved not acting on the fact that all instruments were recalibrated on 6/13/95. The test apparatus used was retired from future use. Surveillance procedures for feedwater temperature instruments and other similar instruments used in the thermal power calculation will be reviewed and revised as necessary. Appropriate personnel involved with the review of out-of-tolerance reports for test apparatus were informed of this event. Personnel involved with the review of completed surveillances will be informed of this event. The programs for reviewing test equipment failure trends and test equipment out-of-tolerance reports will be reviewed. There were no previous similar events identified.
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TEXT (M more spece a requend, une annonel NRC Fom, kl6A's) (1h Reauirements of the Report This report is submitted pursuant to 10 CFR 50.73(a)(2)(i)(B) due to a condition prohibited by Tcchnical Specifications (Tech Specs) due to a slight increase above rated thermal power.
Tcch Spec 1.1 requires that rated thermal power shall be a total reactor core heat transfer rate to the reactor coolant of 3458 Megawatts thermal (MWt).
Unit Cond;tions at Time of Discovery and Occurrence of the Event At the time of discovery of the event, Unit 3 was at 100 percent indicated reactor power. The cvent occurred when Unit 3 achieved 100 percent indicated reactor power on 10/22/95 after r turning from a refueling outage.
There were no other systems, structures, t components that were inoperable that contributed to the event.
Description of the Event On 1/21/97, at 1450 hours0.0168 days <br />0.403 hours <br />0.0024 weeks <br />5.51725e-4 months <br />, licensed operating personnel determined that the plant thermal power level exceeded 100 percent by approximately 0.6 percent. This was based on the completion of a routine surveillance check that resulted in recalibration of all four feedwater t:mperature instruments (Ells:T). These instruments provide input into the overall core thermal power calculation. The recalibration changed the feedwater temperature indication by cpproximately five degrees F, resulting in an indicated thermal power increase from 3450 to 3479 MWt. Once identified, operating personnel promptly reduced reactor power to within limits.
investigation into this issue, identified that the four feedwater temperature instruments had been r: calibrated on the previous routine surveillance check completed 6/13/95. At this time, Unit 3 was in Mode 1 (RUN) operating at 82 percent reactor power and coasting down for a refueling outage. The calibration resulted in the adjustment of all four instruments in a non-conservative direction.
Cause of the Event
The non-conservative calibration of the feedwater temperature instruments on 6/13/95 was due to the use of out-of-tolerance test equipment. This test equipment was determined to be out-of-tolerance during a check of this equipment in July 1996. This check was performed to ensure that the test equipment was performing properly and initiated actions to ensure that equipment calibrated with the out-of-tolerance test equipment was evaluated.
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When the out-of-tolerance test equipment was reported to station personnel in July 1996, no r: calibrations were performed on the feedwater temperature instruments that had previously been calibrated with the out-of-tolerance test equipment. This was due to a lack of specific informatie.,n concerning the function of the feedwater temperature instruments in the surveillance t:st tnt calibrates the instruments. The surveillance test was not c! ear that these instruments are i. sed as inputs to the thermal power calculation. Because of this, Instrumentation & Control (l&C) personnel (utility, non-licensed) reviewing the out-of-tolerance report and the surveillance precedures did not recognize the impact on the thermal power calculation. These feedwater tmperature instruments are not used for other important to safety functions other than the thermal power calculation.
i Subsequent investigation has revealed that the test equipment has had a history of internal compor'ent failures although the failures were on different components within the test equipment. However, this overall trend of problems did not result in the retirement of the equipment. Other similar test equipment was found to have generally good reliability.
1 A contributing cause to this event occurred when the calibrations were performed on 6/13/95 during plant coastdown for a refueling outage. At this time, personnel did not act on the fact that the surveillance check resulted in all four instruments being recalibrated. For routine surveillances, it would be unusual that all instruments were in need of calibration. The current
. practice is that all four instruments are calibrated within a short span of time and are reviewed as part of one document. Additionally, the calibrations on 6/13/95 were performed subsequent to a plant power change evolution. Therefore, power changes as a result of the recalibrations were not immediately obvious due to stabilization of plant power subsequent to the power maneuver.
Analysis of the Event
There were no actual safety consequences to the plant. The miscalibration of the feedwater t::mperature instruments resulted in thermal power being 0.6 percent greater than indicated.
This deviation is well within design basis event assumptions which account for thermal power excursions up to 102 percent. Had a design basis event occurred during the time that power was 0.6 percent greater than allowed, there would have been no impact on the ability to mitigate the event. Additionally, this deviation has been determined to have no impact on balance of plant or reactor systems.
Because the deviation was introduced during a time when the plant was in coastdown for the Unit 3 refueling outage in September 1995, there were no actual violations of thermal power 0$Cferm3 sea (649)
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Corrective Actions
Reactor power was immediately reduced to 100 percent thermal power upon discovery of the event on 1/21/97. Unit 2 was determined to be operating within limits.
l The test apparatus used was retired from future use. Other similar test instruments have been ovaluated and found acceptable for continued use. Equipment calibrated with the retired test equipment was reviewed and verified to be acceptable for continued operation. Other similar l
tasting on instruments that input plant power calculations were reviewed and found to not have impacted the plant.
I Surveillance procedures for feedwater temperature instruments and other similar instruments I
used in the thermal power calculation will be reviewed and revised as necessary.
P Appropriate personnel involved with the review of out-of-tolerance reports for test apparatus 1
were informed of this event.
Personnel who review completed surveillances will be informed of this event involving the need l
to maintain a questioning attitude when reviewing completed test results.
I The program for reviewing test equipment failure trends will be evaluated and revised as n:cessary, i
The program for maintenance and test equipment out-of-tolerance reports will be reviewed and revised as necessary to ensure that adequate controls exist for the review of these reports.
l
Previous Similar Events
There were no previous similar LERs identified involving a non-compliance with the rated l
thermal power limit due to the use of inaccurate test equipment. There was a previous event i
(LER 2-96-01) involving the exclusion of the Control Rod Drive flow element into the overall core thermal power calculation. However, the corrective actions in LER 2-96-01 involved ensuring the correct inputs into the calculation and, therefore, would not have prevented this cvent. The errors introduced in that event were very small (approximately 1 MWt) and do not 1
impact the safety analysis in this report.
N2C Fece 306A (6 891 1